ML18038A884

Provides Revised Relief Request SPT-4 for NRC Review

ML18038A884
Person / Time
Site:	Browns Ferry
Issue date:	09/01/1994
From:	Salas P TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9409120307
Download: ML18038A884 (20)

Text

5'H.IC3H.j.MW ACCELERATED RIDS PROCESSING) 1 REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9409120307 DOC.DATE: 94/09/01 NOTARIZED: 'NO DOCKET g FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH. NAME AUTHOR AFFILIATION SALAS,P. Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Provides revised relief request SPT-4 for NRC review.

DISTRIBUTION CODE: A047D TITLE: OR COPIES RECEIVED:LTR Submittal: Inservice/Testing/Relief from L ENCL L SIZE:

ASME Code GL-89-04.

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-4 1 0 PD2-4-PD 1 1 WILLIAMS,J. 2 2 INTERNAL: ACRS 6 6 AEOD/SPD/RAB 1 1 NRR/DE/EMCB 1 1 NRR/DE/EMEB 1 1 NUDOCS-ABSTRACT 1 1 OC/~LCB 1 0 OGC/HDS3, 1 0 RE~LE 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: EG&G BROWN, B 1 1 EG&G RANSOME,C 1 1 NOAC 1 1 NRC PDR 1 1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 504-2083 ) TO ELIMINATEYOUR NAME FROM DISTRIBUTIONLISTS FOR DOCUMENTS YOU DON'T NEEDl TOTAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 19

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Tennessee Valley Authority, Post Office Box 2000. Decatur. Alabama 36609 10 CFR 50.55a(3)(ii)

September 01, 1994 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of ) Docket No. 50-260 Tennessee Valley Authority )

BROWNS FERRY NUCLEAR PLANT (BFN) AMERZCAN SOCZETY OF MECHANZCAL ENGZNEERS (ASME) SECTZON ZZ ZNSERVZCE SYSTEM PRESSURE TEST PROGRAM REUZSED RELIEF REQUEST SPT-4 In accordance with 10 CFR 50.55a(3)(ii), enclosed is a revised Request for Relief from the specified Section XI systems pressure testing requirements of the 1986 Edition of the ASME Boiler and Pressure Code for NRC review.

TVA identified the impractical aspect of the specified ASME Section XI requirements during the recent preparations for the Unit 2, Cycle 7 refueling outage. With Unit 2 scheduled to begin this outage as early as October 1, 1994, we request, expeditious review and approval of the attached Request for Relief.

MOJO 8 94091203'07 940901 PDR ADOt.K OS000260

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I U. S. Nuclear Regulatory Commission Page 2 September '01, 1994 There are no commitments contained in this letter. If you have any questions, please contact me at (205) 729-2636.

Sincerely, Pe o Salas Manager of Site Licensing Enclosure cc: see page 3

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U.S. Nuclear Regulatory Commission Page 3 September 01, 1994 cc (Enclosure):

Mr. Mark S. Lesser, Section Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637 Athens, Alabama 35611 Mr.. J. F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852

Ii 0 ENCLOSURE TENNESSEE VALLEY AUTHOR1TY BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 2 REVISED REQUEST FOR REL1EF SPT-4 System: Control Rod Drive (CRD) Hydraulic (85)

Drawing: 2-47E820-2 Components: Control rod drive cap screws (185 CRDs per unit, 8 cap screws per CRD housing;to-flange connection)

Class:

Function: Connects CRD to the reactor pressure vessel (RPV)

CRD nozzle flange Impractical Test Requirements: IWA-5250(a) The source of leakage detected during the conduct of a system pressure test shall be located and evaluated by the Owner for corrective measures as follows:

"(2) if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100."

Basis for Relief: In accordance with the requirements of Table IWB-2500-1, Examinati'on Category B-P, Item No.

B15.10, a leakage test of the reactor pressure vessel pressure retaining boundary is conducted prior to plant startup following each reactor refueling outage. The leakage test is conducted at nominal system pressure (1005 psig at the RPV dome) immediately prior to the startup of the unit. This examination includes the 185 CRD connections located on the bottom of the reactor pressure vessel. During re-pressurization following unit refueling, it is not uncommon to have small amounts of leakage at some of the CRD connections.

Compliance with subparagraph IWA-5250(a) (2) in the event of leakage at a CRD connection would result in an extreme hardship which is not commensurate with the increased level of safety that would be achieved. The hardship is due to the requirement that for any CRD connectors where leakage is detected during the pressure test the connector cap screws must be removed and examined. This requires that the RPV be depressurized and the CRD restraint be removed to permit removal of the CRD bolting.

The Nuclear Steam Supply System (NSSS) supplier, General Electric (GE), has informed Boiling Water Reactor (BWR) owners that leakage from these cap

.screw connections is a common occurrence, and in most instances leakage stops within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the connection being pressurized to 1000 psig.

Industry experience as documented in GE Nuclear Energy Services Information Letter (SIL) No. 483, Revision 2, dated August 5, 1992, has shown that these CRD cap screws are susceptible to stress corrosion cracking. Due to this susceptibility, GE has recommended the replacement of these cap screws with a new design and higher strength material cap screw, and a new design washer to facilitate drainage.

This new design is being incorporated at Browns Ferry. The cap screws for 26 CRD units were changed during the Unit 2 Cycle 6 refueling outage. Fourteen CRD units, including the cap screws not previously changed to the new bolting design, are scheduled to be changed during the Unit 2 Cycle 7 refueling outage.

The remaining cap screws are normally changed as the CRDs are replaced. 'The cap screws, which are removed, are examined in accordance with Section XI of the ASME Boiler and Pressure vessel code.

None of the 208 cap screws thus far examined have exhibited indications of stress corrosion cracking.

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0 As documented in SIL 483, GE has determined that based on the evaluation of crack data, structural integrity and plant safety are not effected by this situation. This evaluation is based in part on the following:

1. three uniformly distributed uncracked cap screws are capable of supporting the CRD loads, and'he probability that through-wall cracks will occur in five or more cap screws on a single CRD is extremely small;

2. if such a failure were to occur, leakage at the connection would precede failure, and the leak detection system and drywell'emperature monitoring system would detect this leakages

3. the CRD support structure under the reactor vessel would allow the 'CRD to drop a maximum of one inch in the event of bolted joint failure;

4. the evaluation of the loss of one CRD has been considered in the plant safety analysis report.

By TVA letter to,NRC, dated April 8, 1993 (Reference 1), TVA stated that all leakage from the CRD connections would be documented and evaluated based on the GE recommendations during the Class 1 component leakage test following refueling during the Cycle 6 outage.

The, Cycle 6 inspection showed 36 CRDs were initially leaking during the .RPV Operational Leak Test. Three of these were leaking at a rate in excess of 30 drops per minute (DPM). Maintenance is normally recommended for leaks that are greater than 30 DPM, which do not exhibit a decreasing trend. Two of the CRDs quickly showed a decrease to less than 30 DPM and the remaining CRD showed a leakage of approximately 40 DPM with a decreasing trend. These leak rates were evaluated by TVA and GE and determined to be acceptable.

In the April 8, 1993 letter, TVA requested the VT-3 examination of all eight cap screws at the CRD connections, where leakage is detected during the leakage test, be deferred to the next refueling outage. This request was modified on

4l May 5, 1993 (Reference 2) to only address Unit 2.

The request was approved for Unit 2 in NRC letter to TVA, dated May 21, 1993 (Reference 3).

TVA has performed additional reviews of this issue since the original relief request and has identified additional pertinent information.

These reviews indicate that bolt failures have primarily occurred in pressurized water reactors (PWRs), at both ambient and elevated temperature environments. The following three causes of bolting failures have been identified and have been evaluated for impact at BFN:

1. 'Stress Corrosion Cracking: This mechanism requires a wet or humid environment, high preload stresses, use of lubricants containing molybdenum disulfide, and/or improper heat treatment of material.

2. Fatigue: This failure is primarily induced by improper preload torquing.

3. Borated Water: This failure is caused by chemical attack from borated water leakage.

These failures have a low probability of occurrence on the CRD cap screws at BFN for the reasons provided below:

1. Approved lubricants are used at BFN and are controlled by procedures. The primary lubricant for this application is Never-Seez, which does not contain molybdenum disulfide.

In addition, the atmosphere in the drywell is required by Technical Specifications to be inerted with nitrogen during power operations. This would deprive the CRD cap screws of free oxygen that would aid in chemical and stress corrosion cracking.

2. The CRD cap screws at BFN are torqued under administrative controls to 350 foot pounds, which results in a preload stress of less than 33 percent of the yield strength.

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'3. Unlike pressurized water reactors, BFN does not use borated water in its primary coolant system for reactivity control during normal operating conditions. The reactor coolant E-4

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system uses demineralized water and is monitored for chemical composition and contaminants.

BFN Technical Specification 3.6.C.1 requires unidentified reactor coolant leakage into primary containment exceed '5 gpm. In addition, total reactor coolant system leakage into -primary containment is, prohibited from exceeding 25 gpm.

The average Unit 2 Cycles 6 and 7 unidentified and identified leakage rates have been well below these limits.

The relatively small increase in safety that could be attributed to the performance of a VT-3 examination of all eight cap screws at the 36 CRD connections, where leakage was detected during the previous leakage test, is offset by the increase in personnel dose exposure and cost that are associated with these examinations.

In summary, TVA requests revised Relief Request SPT-4 be approved for the following reasons:

1. TVA's ongoing incorporation of GE recommended revised CRD cap screw design that is more resistant to stress corrosion cracking.

2. None of the cap screws thus far examined have exhibited indications of stress corrosion cracking.

3. GE has determined that based on the evaluation of crack data, structural integrity and plant safety are not effected by this situation.

4. The CRD leakage identified during the previous outage was evaluated and determined to be acceptable.

5. Bolting failure mechanisms have been evaluated and determined to have a low probability of occurrence.

6. The relatively low average unidentified and identified leakage rates during Unit 2 Cycles 6 and 7.

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7. The relatively small increase in safety that could be attributed to the performance of a VT-3 examination of the CRD connections is offset by the increase in personnel dose exposure and cost that are associated with these examinations.

Alternate Testing: A VT-1 inspection will be performed on cap screws from CRD connections that are disassembled. In addition, a fluorescent magnetic particle surface examination will be performed in accordance with ASME Section XI and GE SIL No. 483 if determined necessary by visual examination. This the VT-1 examination will provide an acceptable level of quality and safety in that the sample of bolts examined will be sufficient to identify degradation trends that do occur.

References:

1. TVA letter to NRC, dated April 8, 1993, American Society of Mechanical Engineers (ASME)Section XI Inservice System Pressure Test Program

2. TVA letter to NRC, dated May 5, 1993, Units 1 and 3 Withdrawal of Control Rod Drive (CRD)

Request for Relief from the American Society of Mechanical Engineers (ASME)Section XI Inservice System Pressure Test Program

3. NRC letter to TVA, dated May 21, 1993, Safety Evaluation of Requests for Relief from American Society of Mechanical Engineers Boiler and Pressure Vessel Code Requirements E-6

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