ML18012A469

Submits Revised Request for Addl Info Re ASME Boiler & Pressure Vessel Code,Section XI ISI Program Relief Request for Regenerative Heat Exchanger

ML18012A469
Person / Time
Site:	Harris
Issue date:	01/21/1997
From:	Robinson W CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-97-004, HNP-97-4, NUDOCS 9701270131
Download: ML18012A469 (12)

Text

REGULAT CATEGORY 1~

INFORMATION DISTRIBUTIONWFSTEM (RIDE)

ACCESSION NBR:.9701270131 DOC.DATE: 97/01/21 NOTARIZED: NO DOCKET I FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION ROBINSON,N.R. Carolina Power 6 Light Co.

RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Submits revised request for addi info re ASME Boiler &

Pressure Vessel Code,Section XI ISI program relief request for regenerative heat exchanger.

DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code GL-89-04 NOTES:Application for permit renewal filed. 05000400 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 LA 1 1 PD2-1 PD 1 1 LE,N 1 1 INTERNAL: ACRS 1 1 AEOD/SPD/RAB 1 1 ENTER 0 1 1 NRR/DE/EMEB 1 1 NUDOCS- BSTRACT 1 1 OGC/HDS2 1 0 RES/DET/EIB 1 1 RES/DET/EMMEB 1 1 EXTERNAL: LITCO ANDERSON 1 1 NOAC 1 1 NRC PDR 1 1 D

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PLEASE HELP US TO REDUCE WASTEI CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED/

TOTAL NUMBER OF COPIES REQUIRED: LTTR 14 ENCL 13

Carolina Power & Light Company William R. Robinson PO Box 165 Vice President New Hill NC 27562 Harris Nuclear Plant SERIAL: HNP-97-004 JAN 21 1997 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 ASME BOILER AND PRESSURE VESSEL CODE, SECTION XI INSERVICE INSPECTION PROGRAM RELIEF REQUEST REGENERATIVE HEAT EXCHANGER REQUEST FOR ADDITIONALINFORMATION

Reference:

Carolina Power & Light Company (CP&L) Letter, Serial Number: HNP-96-162, Inservice Inspection Program Relief Request Regenerative Heat Exchanger, dated October 8, 1996.

Dear Sir or Madam:

In the above-referenced letter, CP&L requested relief from American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI inspection requirements related to the regenerative heat exchanger. On December 10, 1996, a conference call between the NRC and CP&L was conducted to discuss NRC review questions in the following areas:

The method by which normal system leakage is detected The ability to isolate the heat exchanger in the event of leakage Details concerning heat exchanger manufacturing (materials)

Operating, design, and test pressure information Information related to the application of leak-before-break on the shell side To address the NRC request for additional information, CP&L revised the detailed relief requests (R2-006 and R2-007) provided as enclosures 1 and 2, respectively, in the above-referenced CP&L letter. These revised relief requests incorporate the additional information requested by the NRC and are provided as enclosures 1 and 2 to this letter. These requests supersede the initial October 8, 1996 request.

970i270131 'P70i2i PDR ADQCK~05000400 8 '

PDR State Road1134 NewHill NC Tel 919362-2502 Fax 919362-2095

Document Control Desk HNP-97-004/ Page 2 NRC approval is requested by March 1, 1997 to support planning activities for the final inservice inspection period of the first 10-year interval to be performed during the next refueling outage, currently scheduled to begin in April 1997.

Please refer any questions regarding this subject to Ms. D. B. Alexander at (919) 362-3190.

Sincerely, W. R. Robinson KWS/kws

Enclosures:

1. ISI Relief Request No. R2-006

2. ISI Relief Request No. R2-007 c: Mr. J. B. Brady, NRC Sr. Resident Inspector Mr. N. B. Le, NRC Project Manager Mr. L. A. Reyes, NRC Regional Administrator

Document Control Desk HNP-97-004/ Page 3 t

bc: Ms. P. B. Brannan Mr. R. D. Martin Mr. B.H. Clark Mr. W. S. Orser Mr. H. K. Chernoff (RNP) Mr. G. A. Rolfson Mr. G. W. Davis Mr. R. S. Stancil Mr. J. W. Donahue Mr. M. A. Turkal (BNP)

Ms. S. F. Flynn Mr. T. D. Walt Mr. H. W. Habermeyer, Jr. Nuclear Records Mr. G.D. Hicks (BNP) File: HVA-2D Mr. M. D. Hill Mr. W. J. Hindman Ms. W. C. Langston (PE&RAS File)

Enclosure 1 to Serial: HNP-97-004 Shearon Harris Nuclear Power Plant Docket No. 50-400/ License No. NPF-63 ASME Boiler and Pressure Vessel Code,Section XI Request for Relief from Inservice Inspection Requirements No. R2-006 Component: Regenerative Heat Exchanger Construction Code: ASME Section III, 1974 Edition including addenda through the Winter of 1976.

Inspection Code: ASME Section XI, 1983 Edition including addenda through the Summer of 1983.

Code Requirement: Volumetric Examination of the Regenerative Heat Exchanger Head to Shell Welds (2) and the Regenerative Heat Exchanger Tubesheet to Shell Welds (2) in accordance with Table IWC-2500-1, Examination Category C-A, Item No's. C1.20 and C1.30.

Head to Shell Welds: II-RHX-01CHTSW-03 and II-RHX-01CHTSW-06 Tube to Tubesheet Welds: II-RHX-01TSTSW-09 and II-RHX-01TSTSW-12 Basis for Relief: The Regenerative Heat Exchanger, located in the reactor containment building, is part of the Chemical and Volume Control System (CVCS),

and it is in service during normal plant operation. The primary function of the Regenerative Heat Exchanger is to recover heat from Reactor Coolant System (RCS) letdown flow and transfer it to the RCS charging flow.

RCS letdown water flows through the shell side of the heat exchanger while charging water flows through the heat exchanger tubes.

The Regenerative Heat Exchanger consists of three horizontal shell and U-tube heat exchangers that are connected in series and mounted one above the other. The heat exchanger assembly is constructed of austenitic stainless steel and is of welded construction.

The normal operating pressure of the shell side of the regenerative heat exchanger is maintained between 2220 and 2250 psig. The design pressure of the heat exchanger is 2485 psig. The pre-service test pressure of the heat exchanger was 4905 psig.

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to Serial: HNP-97-004

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Enclosure 1 Each individual heat exchanger has four welds that require inspection in accordance with Table IWC-2500-1, Examination Category C-A, Item No's. C1.20 and C1.30. The four welds on each heat exchanger (12 welds total) were successfully inspected during the pre-service inspection period with no rejectable indications noted. Additionally, 8 of the 12 heat exchanger welds on two of the three heat exchangers were successfully inservice inspected (4 welds inspected during Refueling Outage Number 02 in 1989 and 4 welds inspected during Refueling Outage Number 03 in 1991) with no rejectable indications noted.

In the unlikely event of regenerative heat exchanger leakage, the regenerative heat exchanger can be isolated from the RCS and the CVCS.

Isolation can be accomplished by the remote manual closure of the normal charging isolation valve (1CS-492), alternate charging isolation valve (1CS-480), the charging line containment isolation valve (1CS-238), the letdown containment isolation valves (1CS-7, 1CS-8, 1CS-9, and 1CS-11) and the normal letdown line isolation valves (1CS-1 and 1CS-2). In addition, the normal charging line isolation valve has a bypass line which willbe isolated from the RCS by four self-actuated check valves (1CS-500, 1CS-497, 1CS-493, and 1CS-774) in the event of leakage.

As identified in NUREG 1061, Volume 1, "Investigation and Evaluation of Stress Corrosion Cracking in Piping of Boiling Water Reactor Plants,"

operating experience and fracture mechanics evaluations indicate that leak-before-break is the most likely mode of piping failure for austenitic stainless steel piping. The leak-before-break failure mode in piping refers to the concept that cracks propagating through the wall of a pipe, by mechanisms such as stress corrosion cracking or by ductile crack extension due to applied loads, will result in a stable through wall crack that can be reliably detected by leakage before catastrophic failure.

Normal reactor coolant inventory monitoring (Technical Specifications Surveillance Requirement 4.4.6.2.1) would provide an indication of Regenerative Heat Exchanger leakage. Net level changes in the pressurizer and volume control tank could be additional indications that the integrity of the Regenerative Heat Exchanger shell has been breached.

Additionally, leakage could be detected by the Containment Airborne Gaseous Radioactivity Monitoring System, the Containment Airborne Particulate Radioactivity Monitoring System, and the Reactor Cavity Sump Level and Flow Monitoring System during normal plant operation.

The Regenerative Heat Exchanger performs no active role (other than acting as a system pressure boundary) during any accident scenario.

Enclosure 1 to Serial: HNP-97-004 During the design basis accident, it is isolated from the emergency core cooling portion of the CVCS by two redundant valves.

The Regenerative Heat Exchanger is located in the reactor containment building in a normally locked-closed high radiation area. Radiation levels in the area of the regenerative heat exchanger typically exceed 2 Rem/hour and estimated contact dose rates are 4 to 6 Rem/hour. The estimated dose to remove/re-install insulation and complete the examinations for the subject welds is estimated to be 10 Person-Rem.

Based upon: (1) the ALARAconcerns involved in performing these examinations, (2) the acceptable results of the preservice and inservice examinations performed to date, (3) the passive safety function of the Regenerative Heat Exchanger, (4) the leakage detection systems available to detect heat exchanger leakage, and (5) the ability to isolate the heat exchanger from the RCS and CVCS in the event of leakage, compliance with the ASME Section XI volumetric examination requirements for the four remaining welds would result in an undue burden without a compensating increase in the level of quality or safety.

Alternate Examination:

A VT-2 visual examination of the subject welds will be performed during system pressure tests scheduled during the 1997 refueling outage. A VT-2 visual examination is defined in IWA-2212 as an examination to locate evidence of leakage from pressure retaining components, or abnormal leakage from components with or without leakage collection systems as required during the conduct of system pressure or functional test. In accordance with IWA-5242(a), a VT-2 examination may be performed without the removal of insulation by examining the accessible and exposed surfaces and joints of the insulation. Since the Harris Plant has received approval to use ASME Code Case N-498-1, "Alternative Rules for 10-year System Hydrostatic Testing for Class 1, 2, and 3 SystemsSection XI, Division 1," the system shall be pressurized to nominal operating pressure for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to performing the VT-2 visual examination.

Enclosure 2 to Serial: HNP-97-004 Shearon Harris Nuclear Power Plant Docket No. 50-400/ License No. NPF-63 ASME Boiler and Pressure Vessel Code,Section XI Request for Relief from Inservice Inspection Requirements No. R2-007 Component: Regenerative Heat Exchanger Construction Code: ASME Section III, 1974 Edition including addenda through the Winter of 1976 Inspection Code: ASME Section XI, 1983 Edition including addenda through the Summer of 1983 Code Requirement: Visual (VT-3) Examination of the Regenerative Heat Exchanger Component Supports (6) in accordance with Table IWF-2500-1, Examination Category F-C, Item No F3.10.

Support No's: II-RHX-01H-001 II-RHX-01H-002 II-RHX-01H-003 II-RHX-01H-004 II-RHX-01H-005 II-RHX-01H-006 Basis for Relief: The Regenerative Heat Exchanger, located in the reactor containment building, is part of the Chemical and Volume Control System (CVCS),

and it is in service during normal plant operation. The Regenerative Heat Exchanger consists of three horizontal shell and U-tube heat exchangers that are connected in series and mounted one above the other. The heat exchanger assembly is constructed of austenitic stainless steel and is of welded construction.

The Regenerative Heat Exchanger unit contains six wall mounted supports (two per heat exchanger) that require inspection in accordance with Table IWF-2500-1, Examination Category F-C, Item No F3.10. The six support units were visually examined (VT-3) "on November 29, 1986 during the pre-service inspection period with no reportable indications noted.

Additionally, two of the support units were inspected, one on November 19, 1989 and the other on April 26, 1991, following their removal to support the volumetric inservice inspection of the head to shell welds on two of the three heat exchangers. Again, no reportable indications were noted.

The normal operating pressure of the shell side of the regenerative heat exchanger is maintained between 2220 and 2250 psig. The design

Enclosure 2 to Serial: HNP-97-004 pressure of the heat exchanger is 2485 psig. The pre-service test pressure of the heat exchanger was 4905 psig.

The primary function of the Regenerative Heat Exchanger is to recover heat from Reactor Coolant System (RCS) letdown flow and transfer it to the RCS charging flow. RCS letdown water flows through the shell side of the heat exchanger while charging water flows through the heat exchanger tubes. The heat exchanger and large portions of the component support units are covered with insulation to optimize heat transfer while minimizing heat loss to the local area. The insulation would have to be removed to perform a VT-3 visual inspection of the support units.

The Regenerative Heat Exchanger is located in the reactor containment building in a normally locked-closed high radiation area. Dose rates in the general area of the regenerative heat exchanger typically exceed 2 Rem/hour and estimated contact dose rates are 4 to 6 Rem/hour. The estimated dose to complete the examinations for the six supports (exclusive of a 4 Person-Rem dose for removaVre-installation of insulation) is 3 Person-Rem.

Based upon the ALARAconcerns involved in performing these examinations, the acceptable results of the pre-service examinations performed to date and the passive safety function of the regenerative heat exchanger, compliance with the code visual examination requirements for the six Regenerative Heat Exchanger support units is considered an undue burden and will not result in a compensating increase in safety.

Alternate Examinations:

A remote visual inspection of the Regenerative Heat Exchanger and support units with the insulation in place willbe performed during the 1997 refueling outage.

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