ML20090M819

Provides Revised Response to Bulletin 88-008 Re Thermal Stresses in Piping Connected to Rcs.Nonproprietary WCAP-12388 & WCAP-13245 & Proprietary WCAP-12387 & WCAP-12425 Repts Encl.Proprietary Repts Withheld

ML20090M819
Person / Time
Site:	Byron, Braidwood
Issue date:	03/16/1992
From:	Chrzanowski D COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML20034D376	List: ML20090M819 ML20090M823 ML20090M828 ML20090M834 ML20090M839
References
IEB-88-008, IEB-88-8, NUDOCS 9203250320
Download: ML20090M819 (17)

Text

. - _ . . _- - . ~ .

, ._ _ Commonwealth Edisori j 14(9 Opus Place s Downers Grove. Illinois 00515 I ,J 4-pq -

March 16,1992 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk

Subject:

Revised Response to Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems.

Byron Units 1 and 2, NRC Docket Numbers 50-454 and 50-455 Braidwood Units 1 and 2, NRC Docket Numbers 50-456 and 50 457

References:

(1) Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems, dated June 22,.1988.

(2) Bulletin 88 08, Supplement 3, dated April 11,1989.

(3) A. Hsla (NRC) to T. Kovach (CECO) letter dated December 11,1991.

(4) " Evaluation Criteria for Responses to NRC Bulletin 88-08, Action 3 and Supplement 3"

_ provided in Reference (5) -

~

(5) Tekconference be_ tween NRC/NRR, CECO and Westinghouse on Thursday February 13,1992.

(6) M.H. Richter to NRC letter dated July 17,1989

Dear Dr. Murley:

! The purpose of this letter is to provide a revised response to Bulletin 88-08

- for Byron and Braidwood Stations. Because of clarifications provided by NRC staff in Reference (3) and in the Raference (5) phone call, Commonwealth Edison (CECO) is ravising the Byron and Braidwood Bulletin 88 08 programs to include temperature >

monitoring provisions. Also, based on CECO's review of the Reference (4) criteria recently provided by the NRC, a previously identified location is no longer considered ~

susceptible to thermal fatigue cracking and is no longer in the scope of Bulletin 88-08 actions for Byron'and Braidwood Stations.

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Dr. Thomas E. Murley March 10,1992 After completion of the temperature monkoring installation, the Byron and Braidwood Bulletin 88-08 programs will piovide for the monitoring of potentially susceptible locations. A discussion of; Bulletin requirements, affected areas. the analysos performed, as well as the monitoring and examination programs is contained in Attachment (1). Attachment (2) is a discussion of industry efforts for final resolution of Bulletin 88-08.

Enclosed in Attachment 3 are:

1. 1 Copy of WCAP-1287,

• Evaluation of Thermal Stratificat6n for the Byron and Braidwood Units 1 and 2 Residual Heat Removal Lines" r (Proprietary).

2. 1 Copy of WCAP-12388, ' Evaluation of Thermal Stratification for the Byron and Braidwood Units t and 2 Residual Heat Removal Lines" (Non-Priority).

Also enciosed in Attachment 3 are a Westinghouse authorization letter, CAW 92-277, accompanying affidavit, Proprietary Informatica Notice, and Copyright Notice.

Enclosed in Attachment 4 are:

3. 1 Copy of WCAP-12425," Evaluation of Byron and Braidwood Units 1 and 2 Auxiliary Spray Lines per NRC Bulletin 88 08"(Proprietary).

4. 1 Copy of WCAP-13245," Evaluation of Byron and Braidwood Units 1 and 2 Auxiliary Spray Lines per NRC Bulletin 88-08"(Non Proprietary).

Also enclosed in Attachment 4 are a Westinghouse authorization letter, CAW-92-278, accompanying affidavit, Proprietary infonoation Notice, and Copyright Notice.

As items 1 and 3 contain information proprietary to Westinghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the Commission's regulations.

Accordingly, it is respectfully requested that tha information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.700 ol the Commission's regulations.

/scl:1619:2

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Dr. Thomas E. Murley Ma*ch 16,1992 Correspondence with respect to the copyright or proprietary aspects of the house Affidavit should reference items listed and/or CAW 92-277 aboveCAW-92 or the supporting 278 and shouWestino'Id be addressed to Nicholas J. Liparuto' Manager of Nuclear Safety and Regulatory Activities, Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

To the best of my knowledge and belief, the statements contained in this document are true and correct, in some respect these statements are not based on my personci knowledge, but on information f urnished by other CECO employees, contractor employees, and consultants. Such information has been reviewed in

. accordance with company practico, and I believe it to be reliable.

If there are any questions or comments, please contact me at (708) 515-7292.

Sincerely, J

4 w $

David J. Chrzanowski .

~

Nuclear Licensing Administrator Generic issues a

4 cc: 'A. Bert' Davis, Regional Administrator-Rill (w/o Attachments 3 and 4)

R. Pulsifer, Pro]ect Manager-NRR/PDill-2 A..Hsia, Project Manager-NRR/PDill 2 (w/c Attachrnents 3 and 4)

R. Elliott,' Project Engineer NRR/PDill-2 (w/o Attachments 3 and 4)

S. DuPont, Senior Resident inspector (Braidwood) (w/o Attachments 3 and 4) -

W. Kropp, Senior Rasident inspector (Byron) (w/o Attachments 3 and 4)

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- /scli1619:3

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ATTACHMENT 1 DULLEllN 88-08 REQUIREMENTS AND COMMONWEALTH EDISON PROGRAM DESCRiPTlON

Background

Reference 1) roquested that licensees take actions to assure that certain reactor coolant system RCS) lines will not be subjected to unacceptable thermal stresses. The first of these ac ons requested !!censees review the systems connected to the RCS to determine if unisolable sections of piping could be subjected to stresses from tem aerature stratification or te'operature oscillat,ons. If these connecting lines could not :>e subjected to this type of condition, no additional actions were required. Action (2) of the E ulletin requested licensees to nondestructively examine the welds, heat affected zones, and high stress areas of those lines identified in Action (1) to assure that there weru no existing flaws. Finally, Action (3) requested licensees to plan and implement a program to provide continuing assu7ance that the piping, identified in Action (1), will not be subjected to combined cyclic ar d static thermal stresses that could restdt in fatigue failure.

Referance (2) notified licencees of another instance of thermally induced fatigue eracking and subsequent RCS leakage. This incident at the Genkal Plant in Japan involved the Residual Heat Removal (RHR) line. The supplemant did not require any additional actions but rec uested that the RHR lines be evaluated for thermal faUgue cracking susceptibility. C ECo responded to this Supplement in Roference (6) and included the RHR pump suction I nes of Byron and Braidwood into the Dultetin 88 08 program.

- Aftoctedlocations Originally the CECO Bulletin 88 08 program had identified, and previous submittals had discussed, three system connections as part of Bulletin 88 00 Action (1). These connections were; safety injection charging line to RCS (four lines per unit), the auxiliary pressuilzer spray (aux spray) inne to main pressurizer spray line (one line per unit), anc the RHR pump suction line to RCS (two locations per unit).

However, after reviewing Reference (4), CECO has determined : hat the aux spray line to main spray line is conhgured in a way ilmt precludes unacceptable therrnal stresses from developing. That is, the aux spray check valves are located approximately 50 to 52 pipe diameters from the main spray r azzle at all Byron and Braiowood units (see Figures 1 through 4). This is much greater than the 25 pipe diameter distance defined in the NRC ovaluation criteria section 3.1(D)(b) as exempting lines from Bulletin considerallon. At this dis'ance any cooler water forced upward from the lower section trap by potential check va!v9 leakage would be warmed to main spray temperature before reachir;g the nozzle.

/scl:1619:4

ATTACHMENT 1 (continued)

CECO Bulletirt0E06_Engineerlog Analyses As explained in the Reference (5) conversation, the CECO Bulletin 88-08 program is not simply an ISIinspection based proaram, it is a program based on engineering analyses that concluded that for two of the Byron and Braidwood locations ilis unlikely that the thermal fatigue phenomenon described in the Bulletin could occur. However, rather than remove the aux spray lines and RHR lines from consideration, CECO elected to perform temporary monitoring and commission additional analyses to determine the impact of worst caso leakage.

To address the continuing assurance provision of Bulletin 88 08 for the auxiliary spray line. CECO had instrumented the Byro;) Units 1 and 2 auxiliary spray lines with surface-mounted temperature sensors to detect adverse temperature distributions. A review of the monitoring data indicated that the temperatures were steady, with no cycling observed. The temperatures were also within the ex aected range, and it was therefore concluded that inteakage of cold fluid from th9 aux lian/ spray line into the main spray piping was not occurring. To ensure the integrity of the piping over the design life of the plant, analysis was performed assuring valve loakage and associated stress cycling. A conservative transient was developed, based upon the experience of the Farely safety injection piping failure. Transient stresses were calculated and used as input to a fatigue analysis, using approved ASME Section XI methodology, to determine an acceptable period of operation between inservice inspection intervals, The following conservatisms were assumed in the overall approach:

• lsolation valve leaks continuously.

- Top-to-bottom of pipe temperature difference is 300'F, based on comparison of maximum potential temperaturs difference with Farley safety injection.

Cyclic period of 7.3 minutes, based on heat transfer calculations.

Fluid temperature transient is instantaneous, i.e. step change in time.

• Initial crack size of 10% of the wall thickness.

Final crack size limited to 60% of the wall thickness.

The results of this analysis was that 39 months of power operation is an acceptable period for inservico inspection intervals to provide continuing assurance of the pressurizer auxiliary spray piping integrity, ar:d therefore monitoring is not necessary. This infonnation is documented in Westinghouse Report WC AP-12425,

- October 1989 (Attachment 4).'

To address the continuing assurcnce provision of the Bulletin 88 08 for the residual heat removal (RHR) suction piping, analysis was pednrmed, and is documented in Westinghouse Report WCAP-1238'7 (Attachment 3). A comparisnn was first made to the Genkal RHR configuration which support the conclusion that a Genkai-type transient is unlikely to occur at Byron and Braidwood. Specifically, transients are not ex aected since turbulant penetration of hot RCS water is expected to extend nearly to the volation valv6.

/ sci:1S t 9:5

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ATTACHMENT I (continued)

Evaluations were carried out, however, assuming valve leakage and associated thermal transients, based on the experience of the Genkal piping f ailure. Trensient stresses were calculated and used as input to a fatigue analysis. Since ASME fatigue usage tactor requirements could potentially be exceeded under the assumed conservative transient loadings, f atigue crack growth calculations were performed to determine an acceptable inservice inspection interval, The following conservatisms were assumed in the overall approach:

. Entire length of horizontal piping stratified.

. Most stiffiy supported configuration of the eight locations at Byron and Braidwood was analyzed, a Maximum moment stress and maximum through wah stress assumed to occur at any location in the horizontal pipmg.

. Cyclic petiod parametrically investigated from 10 to 60 minu!es.

Initial flaw size of 15% of the wall thickness.

Final flaw size of 60% ef the wall thickness.

This conservative analysis concluded that 4.1 years of continuous valve leakage and transient loading would result in a crack of 60% of the wall thickness, inservice inspection was therefore recommended at every other refusting outage, or about three yearts of operation.

CECO recognized that much was unknown about the Farley, Tihange and Genkai events and therefore had chusen to supplement the analyses with an enhanced, augmented ultranonic examination program. This examination exceeds ASME Section XI requirements in several aspects. First, the frequency of examination is every other refueling. Second, the examination uses IGSCC techniques and EPRI qualified examiners. Also,in the case of the aux spray location, a mock up of the branch connection was developed to optimize the inspection.

However, CECO understands the NRC's concerns regarding fatigue crack initiation and is now proposing a temperature monitoring program for the RHR location. The monitoring program will replace the interim examination program. l Boposeditadhisting Monitodng Plograms The proposed monitoring program for the RHR lines will follow the guidelines and exceedance criteria established in Reference (4). Prior to the installation of the mon!toring equipment at Byron and Braidwood, RCS integrity at the RHR connection will be assured by the existing analyses and examinations described above. Il after monitoring the RHR location it is determined that turbulent penetration (discussed previously) is in f act occurring, CECO will discuss with the NRC the option of discontinu:ng temperature monitoring on RHR.

/scl:1619:6

ATTACHMENT I (continned)

The aux spray locations, as stated previously, will no longer be considered as candidate locations for a Bulletin 88 08 rnonitoring program. The aux spray location will, however, remain in the Bulletin 88-08 ultrasonic examination program until the re=ults of the EPRI program, described in Attachment (2), are available to evaluate this location.

CECO proposes !a maintain the ongoing leakage monitoring program for the four (per unit) safety injection locations as an alternative to temperature monitoring. This leakage monitoring program meets the requirements of Action (3) of the Bulletin which states;" Plan and implomont a program to provide continuing assurance that unisolable sections of all piping connected to the RCS will not be subjected to combined cyclic and static thermal and other stresses that could cause fatigue f ailure during the remaining life of the unit." The CECO leakage monitoring program, by assuring leak tight integrity of the isolation valves, prevents thermal stresses from developing thereby preventing any f atigue crack initiation. A single line drawing, to help explain the monitoring program,is shown in Figure (5). Althouoh the drawing is listed as a Braidwood Unit 1 drawing it is typical of the Byron Onits 1/ 2 and Braidwood Unit 2 configurations.

The leah test procedure which monitors any leakage past the isolation valves was spacifically developed for Bulletin 88-08. This procedure is part of an overall Technical Specification surveillance that also monitors the leakage by the check valves in this portion of the Safety injection system.

Simply described the procedure records the leakage past isolation valves Sl8801 A-1 and Sl88018-2 by monitoring leakane through test connection S1044. First the piping downstream of check valve Sl8815 and the isolation valves is depressurized.

Next, the charging aumps are staded and are run against the closed Sl8801 A-1 and Sl88010 2 valves. t is then that any leakage is accurately measured usiag a graduated cylinder and stopwatch through the test connection S1044. Leakage past theco isolatico valves is required to be reported and reviewed by Engineering for impact on the Bulletin 88-08 program. To date, no leakage has Feen detected passed these isolation valves at Byron or Braidwood. This monitoting program has assured that unacceptable thermal stresses have not been induced into the down stream piping or RCS nozzle connection.

This isolation valve leakage monitoring is pedormed,

a. At least once per 18 months;

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 months; ,

c. Prior to returning the valve to service following maintenance; repair, or replacement work on the valve;

d. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action or flow through the valve, except for valves 1RH8701 A and B and 1RH8702A and B.

/scl:1619:7 i

4 ATTACHMENT t (continued)

This monitoring frequency encompasses all occasions when these isolation valves may have a potential to leak, that is af ter start up, after stroking or aller maintenance.

IndVStryff0 gram Ccmmonwealth Edison is closely following the industry's efforts in optimizing a Bulletin 88-08 program. The ultimate solution to this issue will rely on the results of EPhl/TASCS program described in Attachment (2). CECO plans to review the EPRI results and take appropriate actions when these results become available.

Conclusions To summarize, all locations potentially susceptible to thermal f atigue cracking at Byron and Braidwood are or will be monitored, The monitoring program for the safety injection lines, four per unit, is an ongoing leakage monitoring program and consists of verifying that the isolatien valves upstream of the check valves do not leak. Assuring that these valves do not leak satisfies the Bulletin requirement that these sections of ,

pipe are not subjected to thermal stresses that could cause fatigue failure.

The temperature monitoring equipment for the RHR lines will be installed at Byron and Braidwood as follows:

Unit Outage Braidwood Unit 1 A1RO4 Spring 1994 Braidwood Unit 2 A2RO3 Spring 1993 l Byron Unit 1 B1ROS Spring 1993 Byron Unit 2 B2RO4 Fall 1993 The current Bulletin 88 08 ultrasonic examinations of the RHR high stress locations will continue until the monitoring eouipment is in place. Com aletion of the RHR monitoring installation will satisfy the requiraments of Bulletin 68 08 'or Byron and Braidwood Stations.

l The aaxiliary spray line, because of the location of the eneck valve relative to the main spray connection. is no longer part of the Bulletin 88-08 monitoring program.

! However the high stress locations on these lines will continue to be ultrasonically examined every other refueling, sutage until the EPRI TASCS program is complete. At that time, CECO will determine 1, .ny locations should be added to or deleted from the Bulletin 88 08 monitoring and examination program.

/ sci:1019:8

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< l ATTACHMENT 2 INDUSTRY EFFORTS FOR RULLETIN 8848 RESOLUTION

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EPRI TASCS PROGRAM The EPRI TASCS (Thermal Stratification, Cycling and Striping) program has been developed to provide industry with the tools needed to evaluate the impact of thermal stratification issues in piping systems.

The need for this program was determined following two years of investigation into tha existing nChods and data to evaluate TASCS phenomena, it was concluded that ihu existing methods were overly conservative and resulted in more monitoring arid inspections that are actually required. This is because most evaluations assumed loadings similar to those which caused the pipe cracks discussed in NRC Bulletin 88-08. Therefore, this program was established to develop more realistic conservative loading for the TASCS phenomena.

The subtasks of the program are as follows:

1. Task 1: Develop Preliminary Categorization and Screening Methodology.

The objectives are: To determine an approach for defining, without extensive analysis, lines affected by TASCS ahanomena; to c'etermine parameters and their ranges to consider in develop ng modes to evaluate TASCS; where possible, establish limits below which thermal fatigue due to TASCS will not occur; and to determine conditions needing experimental support within the project scope.

2. Task 2: Define TASCS Anatytical Approach. The objectives are: To defino a methodology for evaluating susceptible lines that have "f ailed" the screening criteria; to define preliminary inputs for thermal stress and f atigue analysis; and to determine where new correlations and evaluation methodology are required as inputs to later tasks,

3. Task 3: Define Testing and Analysis Program. The objective is to define the scope of testing and analysis based on available data and input from Tasks 1 and 2.

4. Task 4: Conduct Testing and Data Acquisition Analysis. The objective is to conduct testing, data acquisition and analysis as specified in Task 3.

5. Task 6: Develop Correlations /Models. The objective is to develop correlations, models, etc. for screening criteria and evaluation methodology for use by utility operators and design engineers.

6. Task 6: Develop Final Screening and Categorization Methodologies, The

- objective is to finalize the screening and categorization methodo ogy developed in Task 1 using the correlations and methods developed in Task 5.

7. Task 7: Guidance Manual and Evaluation Tools. The objective is to develop guidance manual for determining systems susceptible to TASCS (from Task

6) and for providing engineering tools necessary to evaluate thermal fatigue impact on systems determined to be potentially susceptible.

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4 EPRI TASCS PROGRAM (continumi)

This pro 0 ram is scheduled to be completed in May 1993, at which timo the i guidance manual and screening criteria will be available for utility use. Workshoas are j also p;anned to provide training to the plant operators and engineers who will utii:e  !

this methodology. ,

i Applicability..to NRC_ Bulletin 08:00 The TASCS program is directed by EPRI and a utility advisory committee consisting of rep)resentatives GE . Spacific priorityofhasthebeen major nuclear given owner'sissues to addressing groups (Westinghouse, related to NRC BulletinB&W, CE and 88-08. Therefore testing and analysis has an emphasis on leakage type flows, the potential for this flow to stratify, and the interactions which occur when ihn leakage flow mixes. This program will therefore produce tools directly suitable to evaluating the issues associated with the auxiliary spray line and its interaction with the pressurizer spray line,

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ATTACHMENT 3 1 copy of WCAP 12387 (Proprietary) 1 copy of WCAP 12388 (Non Proprietary)

.. and Westinghouse authorizatico letter, CAW 92-277, accompanying affidavit, Proprietary Information Netico, and Copyright Notico.

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