ML18038A939
| ML18038A939 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/03/1994 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9410120093 | |
| Download: ML18038A939 (20) | |
Text
P M(3RIWY 1 ACCELERATED RIDS PROCESSING)
REGULATORY INFORMATION DISTRIBUTXON SYSTEM (RIDS)
ACCESSION NBR:9410120093 DOC.DATE: 94/10/03 NOTARIZED: NO FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee AUTH.NAME AUTHOR AFFXLIATION SALAS,P.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
DOCKET g
05000260 P
SUBJECT:
Forwards revised Request for Relief SPT-4,providing permanent relief by alternate testing of connections disassembled for routine maint.Response to NRC 940926 telcon RAI re previous relief request also provided.
DISTRIBUTION CODE-A047D COPIES RECEIVED:LTR ENCL SIZE-TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code GL-9-04 NOTES 0
RECXPIENT ID CODE/NAME PD2-4 WILLIAMS,J.
INTERNAL: ACRS~
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1 RECIPIENT ID CODE/NAME PD2-4-PD AEOD/SPD/RAB NRR/DE/EMCB NUDOCS-ABSTRACT OGC/HDS3 EGS(G RANSOME,C NRC PDR COPIES LTTR ENCL 1
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u NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE IIELP US TO REDUCE iVASTE!CONTACTTfiEDOCUMENTCONTROL DESK, ROOhf Pl-37 (EXT. 504-2083 ) TO ELli'ifINATEYOUR NAME PRO.'if DISTRIBUTIONLISTS I'OR DOCI;MENTS YOU DON"I'EED!
TOTAL NUMBER OF COPIES REQUIRED:
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Tennessee Valley Authority. Post Office Box 2000, Decatur, Alabama 35609 10 CFR 50.55a (3)(i)
October 3, 1994 U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter Of Tennessee Valley Authority Docket Nos.
50-260 BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)
SECTION ZI INSERVICE SYSTEM PRESSURE TEST (SPT)
PROGRAM -"'REVISED RELIEF REQUEST SPT-4 In accordance with 10 CFR 50.55a (3) (i), enclosed is a revised Request for Relief from the specified Section XI system pressure testing requirements of the 1986 Edition of the ASME Boiler and Pressure Code for NRC review and approval.
This revised relief request provides permanent relief by alternate testing of connections that are disassembled for routine maintenance.
The revised Request for Relief enclosure 1 supersedes the Request for Relief submitted by BFN on September 1,
1994.
This letter also provides TVA's reply to NRC's September 26, 1994, telephone request for additional information pertaining to the previous relief request.
NRC requested that TVA perform some sampling of the 36 control rod drive (CRD) bolted connections that leaked at the beginning of Unit 2 Cycle 7.
NRC requested information on the status the CRD bolted connections at the end of Unit 2 Cycle 7 operation.
Immediately after the planned shutdown for the refueling
- outage, on October 1,
Enclosure 1 contains the revise request for relief.
The changes are identified with a bar in the right-hand margin.
Enclosure 2 contains the requested information and details of the inspection results.
9'410120093 941003 PDR ADOCK 050002b0 P
I
U.S. Nuclear Regulatory Commission Page 2
October 3, 1994 With Unit, 2 shutdown for refueling and CRD work scheduled to begin within approximately two weeks, TVA requests expeditious review and approval of the attached Request for Relief.
There are no commitments contained in this letter.
If you have any questions, please contact me at (205) 729-2636.
Since l d
Salas Manager of Site Licensing Enclosure cc:
See page 2
U.S. Nuclear Regulatory Commission Page 3
October 3, 1994 Enclosure 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
ENCLOSURE 1
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 REVISED REQUEST FOR RELIEF SPT-4 System:
Drawing:
Components:
Control Rod Drive (CRD) Hydraulic (85) 2-47E820-2 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:
ASME Section XI, 1986 Code Edition, IWA-5250(a)
The source of leakage detected during the conduct of a syst: em 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, Examination 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 syst: em 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.
This leakage is typically on the order of 1-30 drops per minute and 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.
BFN inspection results of the CRDs at the end of the current, cycle (Unit 2 Cycle 7) provide additional evidence to support this information.
Thirty-six bolted connections leaked at the beginning of the Unit 2 Cycle 7 operation.
Only four of the 36 were leaking at the end of Unit 2 Cycle 7 operation.
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 testing and if indications are found, 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 (208 bolts) were changed during the Unit 2 Cycle 6
refueling outage.
Fourteen CRD units, including the new cap screw 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.
E1-2
As documented in SXL 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 the 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 temperature monitoring system would detect this leakage while still at very low leakage rates; 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 from any cause 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
(May 1993) outage.
The Cycle 6
(May 1993) 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 by the vendor (GE) for leaks that are greater than 30 DPM, which do not exhibit a decreasing trend.
Two of the three worst case leaking CRDs quickly showed a
decrease to less than 30 DPM and the remaining worst case leaking CRD showed a leakage of approximately 40 DPM with a decreasing trend.
The other 33 CRDs were well below 30 DPM.
All leak rates were evaluated by TVA and GE and determined to be acceptable.
El-3
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 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 and a drywell inspection of the mating surfaces at pressure 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 procedurally controlled.
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.
El-4
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.
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 system uses demineralized water and is monitored for chemical composition and contaminants.
TVA performed a visual inspection of the CRDs undervessel area immediately after shutdown on October 1,
1994.
The inspection identified five CRDs that were leaking.
No significant leakage was identified.
The worst case leak was two DPM.
The remaining four CRDs were leaking at a rate of one or less DPM.
This inspection was performed at reactor vessel temperature of 533 degrees F and pressure at 920 psi.
Three of the leaking CRDs leaked both Unit 2 Cycles 5 and 6.
Four of the five were among the 36 that leaked during the Unit 2 Cycle 6 test (May 1993).
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:
BFNs CRD maintenance change out program which incorporates 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.
E1-5
4 ~
The CRD leakage identified during the previous outage was evaluated and determined to be acceptable.
Alternate Testing:
5.
6.
7.
Bolting failure mechanisms have been evaluated and determined to have a low probability of occurrence.
Inspection at the beginning of the current outage (Oct.
1, 1994) found no significant leakage.
The VT-1 testing of bolts removed during CRD maintenance and the sampling to be completed during the Unit 2 Cycle 7 outage will provide meaningful sample that provides reasonable assurance that structural integrity and plant safety are not affected.
During the Unit 2 Cycle 7 outage, TVA will perform a VT-1 inspection on the cap screws for the five CRDs that leaked during the October 1,
- 1994, CRD inspection.
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 the VT-1 visual examination.
References:
1 ~
During Unit 2 Cycle 7 and in future outages if the leakage is evaluated and determined to be acceptable no further inspection will be performed.
VT-1 inspections will be performed on cap screws from CRD connections that are disassembled for routine maintenance.
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 the VT-1 visual examination.
This examination will provide an acceptable level of c{uality and safety in that the sample of bolts examined will be sufficient to identify degradation trends should they occur.
TVA letter to NRC, dated April 8,
- 1993, American Society of Mechanical Engineers (ASME)Section XI Inservice System Pressure Test Program E1-6
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 E1-7
ENCLOSURE 2
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2
RESPONSE
TO NRC REQUEST FOR ADDITIONAL ZNFORMATZON REGARDING SYSTEM PRESSURE TEST (SPT)
REVISED REQUEST FOR RELIEF SPT-4 I ~
PURPOSE AND BACKGROUND II
'his enclosure provides TVA's reply to NRC's September 26, 1994, telephone request for additional information regarding control rod drive (CRD) bolting connections.
The following information was requested by NRC at the conclusion of the September 26, 1994 telephone conversation.
NRC Request Provide past history on the 26 (CRDs) that, were replaced during BFN Unit 2 Cycle 6 refueling outage.
TVA's Response During BFN Unit 2 Cycle 6 refueling outage, TVA replaced 26 CRDs.
Even though General Electric (GE) recommends test and replacement of defective cap screws, BFN work practices replace each bolt.
Two of the CRDs that were replaced had leaked during Unit 2 Cycle 5 outage.
The two CRDs (16 bolts) were inspected for corrosion degradation and crack-like flaws and found to be acceptable.
Figure 1 provides additional details on number of leaking CRDs and the replacement of drives per refueling outage.
2.
NRC Request Provide TVA's change out, plans for replacing BWR 4 CRDs with BWR 6 CRDs.
TVA'S Response TVA's plans for CRD change out is in accordance with the CRD maintenance activities.
The maintenance activity involves replacement of remaining BWR 4 devices with BWR 6 drives.
Selections for each outage are based on performance parameters such as stall flows, drive operating temperatures, operational problems (i.e.,
CRD uncoupling) and scram times.
The exact number of drives to be exchanged during each outage is determined based on performance and scheduled considerations, but is expected to be approximately ten percent of the total population of CRDs.
3.
NRC Request Provide the inspection results for CRDs at the beginning of the Unit 2 Cycle 7 outage.
TVA Response TVA performed a visual inspection of the CRDs undervessel area immediately after shutdown on October 1,
1994.
The inspection identified five CRDs that were leaking.
No significant leakage was identified.
The worst case leak was two drops per minute (DPM).
The remaining four CRDs were leaking at a rate of one or less DPM.
This inspection was performed at reactor vessel temperature of 533 degrees F and pressure at 920 psi.
None of the three worst case CRDs which leaked at the beginning of Unit 2 Cycle 7 operation were leaking during the visual inspection immediately after shutdown.
Three of the leaking CRDs leaked both Unit 2 Cycle 5 and 6.
Four of the five were among the 36 that leaked during the Unit 2 Cycle 6 test (May 1993).
4 ~
NRC Request Provide TVA's inspection plans for the CRDs after all (185)
CRDs have been change out.
TVA Response Each outage a reactor vessel and attached piping leak test in accordance with Table IWB-2500-1 examination category B-P, Item No. B-15.10 is performed.
During these tests a
visual test (VT)-2 is performed of the CRD flanges.
5.
NRC Request Provide TVA's plans for sample expansion for the 36 CRDs that leaked during Unit 2 Cycle 6 refueling outage.
TVA's Response TVA identified 36 drives that leaked the last refueling outage (Unit 2 Cycle 6).
The CRD inspection performed at the end of Unit 2 Cycle 7 operation identified five CRDs that leaked.
Four of these five were among the 36.
TVA plans to replace the 0-ring and cap screws on these five CRDs (40 screws).
The 0-ring seals are being replaced as a
preventive maintenance activity to eliminate future leaks.
TVA will perform a VT-1 inspection on the removed cap screws.
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 the VT-1 visual examination.
Unit 2 Cycle 7 normal maintenance will change out one CRD with a history of leakage from Unit 2 Cycle 5 (eight bolts).
E2-2
'his population ensures some of the CRDs that also leaked during the leak test following the Unit 2 Cycle 5 outage are included.
This population of 48 bolts will be inspected (VT-1) for corrosion degradation and crack-like flaws.
During Unit 2 Cycle 6 refueling outage, BFN replaced 26
- drives, two of which leaked during Unit 2 Cycle 5 outage (16 bolts).
This total population of 64 bolts with leakage history provide a statistically representative sample of a random population that provide assurance no corrosion degradation and crack-like flaws exist at a 95/95 confidence level.
See Figure 1 for a summary.
NRC Request Clarify the reference page E-5, 2nd paragraph that states:
"BFN Technical Specification 3.6.C.l requires unidentified reactor coolant leakage into primary containment exceed 5
gpme
~
~
~
II TVA Response The paragraph was deleted based on quantitative results from the preoutage drywell inspection.
Reference TVA letter to NRC, dated September 1,
- 1994, American Society of Mechanical Engineers (ASME)Section XI Inservice System Pressure Test, Program Relief Request SPT-4 E2-3
Figure 1 13 CRD's (104 Bolts)
Beginning of Cycle-7 36 Flange Leaks Beginning of Cycle-6 2 CRD's 32 (16 Bolts) ~ Flange Leaks Cycle 7 Change Bolts on 14 Drives Leaks at end ofU2 C7 Operation Cycle 6 Change Bolts on 26 Drives 1 CRD
'(8 Bolts)
BOC 7 (36)
EOC 7 5
Total CRD Bolting Changed out to be chan ed out 2 (16 Bolts) 1 (8 Bolts) 5 40 Bolts 8 (64 Bolts)
Status Complete