ML050620258
| ML050620258 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/02/2005 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC1704, TAC MC1705, TAC MC1706 | |
| Download: ML050620258 (14) | |
Text
March 2, 2005 10 CFR 54 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen:
In the Matter of
) Docket Nos. 50-259 Tennessee Valley Authority
) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 LICENSE RENEWAL APPLICATION (LRA) - LRA SECTIONS 2.5 and 4.7.8 -
RESPONSE TO NRC REQUEST FOR FOLLOW UP QUESTION FOR RAI B.2.1.5-3, 2.5-2, and 4.7.8 (TAC NOS. MC1704, MC1705, AND MC1706)
By letter dated December 31, 2003, TVA submitted, for NRC review, an application pursuant to 10 CFR 54, to renew the operating licenses for the Browns Ferry Nuclear Plant, Units 1, 2, and 3. As part of its review of TVAs license response letters, the NRC staff, through an informal request on January 31, 2005, identified additional follow up questions for RAI B.2.1.5-3, 2.5-2, and 4.7.8. The questions concentrate on Hydrogen Water Chemistry, Neutron Monitoring Systems, and weld flaw evaluations of EECW piping.
The enclosure to this letter contains the corresponding TVA response to the specific NRC requests for additional information.
U.S. Nuclear Regulatory Commission Page 2 March 2, 2005 If you have any questions regarding this information, please contact Ken Brune, Browns Ferry License Renewal Project Manager, at (423) 751-8421.
I declare under penalty of perjury that the foregoing is true and correct. Executed on this 2nd day of March, 2005.
Sincerely, Original signed by:
T. E. Abney Manager of Licensing and Industry Affairs
Enclosure:
cc: See page 3
U.S. Nuclear Regulatory Commission Page 3 March 2, 2005 Enclosure cc (Enclosure):
State Health Officer Alabama Department of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017 Chairman Limestone County Commission 310 West Washington Street Athens, Alabama 35611 (Via NRC Electronic Distribution)
Enclosure cc (Enclosure):
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 Mr. Stephen J. Cahill, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 NRC Unit 1 Restart Senior Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970 cc: continued page 4
U.S. Nuclear Regulatory Commission Page 4 March 2, 2005 cc: (Enclosure)
Margaret Chernoff, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Yoira K. Diaz-Sanabria, Project Manager U.S. Nuclear Regulatory Commission (MS 011F1)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Ramachandran Subbaratnam, Project Manager U.S. Nuclear Regulatory Commission (MS 011F1)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
U.S. Nuclear Regulatory Commission Page 5 March 2, 2005 DTL:GLS:BAB Enclosure cc (Enclosure):
A. S. Bhatnagar, LP 6-C K. A. Brune, LP 4F-C J. C. Fornicola, LP 6A-C R. G. Jones, NAB 1A-BFN K. L. Krueger, POB 2C-BFN R. F. Marks, Jr., PAB 1A-BFN F. C. Mashburn, BR 4X-C N. M. Moon, LP 6A-C J. R. Rupert, NAB 1F-BFN K. W. Singer, LP 6A-C M. D. Skaggs, PAB 1E-BFN E. J. Vigluicci, ET 11A-K NSRB Support, LP 5M-C EDMS, WT CA-K s://Licensing/Lic/BFN LR Follow Up RAI B.2.1.5-3, 2.5-2, and 4.7.8 TVA Response Letter.doc
ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 LICENSE RENEWAL APPLICATION (LRA),
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAI),
RELATED TO INFORMAL REQUEST FOR FOLLOW UP QUESTIONS FOR RAI B.2.1.5-3, 2.5-2, and 4.7.8 TLAA (SEE ATTACHED)
E-1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2, AND 3 LICENSE RENEWAL APPLICATION (LRA),
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (RAI),
RELATED TO INFORMAL REQUEST FOR FOLLOW UP QUESTIONS FOR RAI B.2.1.5-3, 2.5-2, and 4.7.8 TLAA By letter dated December 31, 2003, TVA submitted, for NRC review, an application pursuant to 10 CFR 54, to renew the operating licenses for the Browns Ferry Nuclear Plant, Units 1, 2, and 3. As part of its review of TVAs license response letters, the NRC staff, through an informal request on January 31, 2005, identified additional follow up questions for RAIs B.2.1.5-3, 2.5-2, and 4.7.8. The questions concentrate on Hydrogen Water Chemistry, Neutron Monitoring Systems, and weld flaw evaluations of EECW piping.
The following contains the corresponding TVA response to the specific NRC requests for additional information.
Follow Up Question to NRC RAI B.2.1.5-3 In response to RAI B.2.1.5-3 the applicant stated that Hydrogen Water Chemistry will be used on Unit 1 following the unit restart with the purpose of reducing dissolved oxygen in the reactor vessel. It is also stated that ECP monitoring is not used at BFN but an alternative model has been used to determine the required hydrogen injection rates. The applicant stated in their response that for the original HWC system in Units 2 and 3 the implementations of noble metals were used for reducing the potential of stress corrosion cracking. Based on your response it is not clear whether noble metals are still added in the system for Units 2 and 3, and which system will be used in Unit 1. Please clarify.
TVA Response to Follow Up Question to NRC RAI B.2.1.5-3 On Unit 2, BFN implemented HWC-M (Moderate - with a 1.6 ppm feedwater dissolved hydrogen concentration target for IGSCC mitigation) in December 1999. In March 2001, BFN Unit 2 applied noble metals (NMCA) which requires a lower feedwater dissolved hydrogen concentration target of 0.25 ppm (HWC-NMCA) for IGSCC mitigation. In August 2000, BFN Unit 3 implemented HWC-NMCA (feedwater dissolved hydrogen concentration target of 0.25 ppm) after applying noble metals (NMCA) in April 2000. It is not unusual for plants to operate for three or four cycles on a single
E-2 noble metal application. Both Units 2 and 3 are currently operating with residual noble metal concentrations from these original noble metal applications. For Unit 1, the current HWC system design (not yet installed) is for HWC-NMCA (feedwater dissolved hydrogen concentration target of 0.25 ppm for IGSCC mitigation).
For all three units and based on future developments, BFN may choose to implement HWC-Moderate, HWC-NMCA, HWC with on-line NMCA (a new yet to be qualified GE process in which even smaller amounts of noble metals are added at some frequency (i.e.
monthly, quarterly, etc.) to the reactor pressure vessel while operating) or other acceptable IGSCC mitigation strategy. Based on BWRVIP-62 and BWRVIP-130, BFN considers HWC-M and HWC-NMCA (and HWC-on-line NMCA when qualified) as acceptable methods of IGSCC mitigation. BFN plans to select (or to change to) the option that offers the best operating strategy for the plant at that time.
Follow Up Question to NRC RAI 2.5-2 (Part A)
The applicant in LRA Section 2.3.3.32, stated that the Neutron Monitoring System (SRM, IRM, LPRM, APRM, OPRM, RBM, and TIP) is in the scope of 10 CFR 54 because it contains components that meet the criteria 10 CFR 54.4 (a)(1) which is safety related.
Furthermore, in response to RAI 2.3.3.32-1, the applicant stated the "spaces approach" was utilized for scoping of electrical components which does not exclude any electrical components from the scope of license renewal. Contrary to the applicant's own statement made in response to RAI 2.3.3.32-1, the applicant, on January 18, 2005, stated that SRM, IRM, and RBM circuits perform no intended function as specified by 10 CFR 54.4 that require them to be included in the scope of license renewal. The applicant also stated that IRMs will generate scram signals during startup conditions for abnormal operational transients (AOT) events, not DBAs. This function is not classified as "safety related." Therefore, cables associated with the SRMs, IRMs and RBMs were screened out and not subject to an AMR. The applicant concluded that only Neutron Monitoring System cables in scope for license renewal are those associated with LPRM.
The staff believes that all cables associated with Neutron Monitoring System should be in the scope of license renewal and cables carrying high voltage and low level signals should be managed by GALL XI.E2 and other instrumentation circuit cables should be managed by XI.E1 program. Provide clarification.
E-3 (Part B)
Furthermore, the applicant on January 18, 2005, stated that medium-voltage cables routed to off-gas treatment building transformer A and B are screened out and not subject to AMR.
The staff agrees with the applicant that the intended functions of off-gas system addressed in LRA Section 2.3.3.19 are accomplished through mechanical means without electrical power.
However, the fans of the standby gas treatment system (LRA Section 2.3.2.2) are in scope and are powered by these transformers. Therefore, the cables should be in scope within standby gas treatment system in LRA Section 2.3.2.2. Based on above, the staff finds that these cables can not be screened out and will require an AMR. Provide clarification.
TVA Response to Follow Up to NRC RAI 2.5-2 Response to Part A In reference to TVAs response to RAI 2.3.3.32-1, the spaces approach to Electrical Scoping and Screening does indeed include all electrical components in the scope of License Renewal, but does not imply that screening cannot occur. TVAs response to NRC follow up to RAI 2.5-2 contained in letter dated January 18, 2004 does not have contradictory statements, it does however state that SRM, IRM, and RBM circuits perform no intended functions that require them to be included in the scope of License Renewal, and therefore are screened out.
The following is taken from BFN Technical Specification Bases Section B3.3.1.1:
Intermediate Range Monitor (IRM) 1.a. Intermediate Range Monitor Neutron Flux - High The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRM) to the lower range of the average power range monitors (APRMs). The IRMs are capable of generating trip signals that can be used to prevent fuel damage resulting from abnormal operating transients (AOT) in the intermediate power range.
Even though the IRMs are capable of generating trip signals, BFN does not credit the IRM Neutron Flux - High scram signal as a safety-related function in the Safe Shutdown Analysis.
E-4 The following is taken from BFN Technical Specification Bases Section B3.3.1.2:
B3.3.1.2 Source Range Monitor (SRM) Instrumentation The SRMs have no safety function and are not assumed to function during any FSAR design basis accident or transient analysis.
The following is taken from BFN Technical Specification Bases Section B3.3.2.1:
B3.3.2.1 Control Rod Block Instrumentation The purpose of the Rod Block Monitor (RBM) is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations.
The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one of the four redundant average power range monitor (APRM) channels supplies a reference signal for one of the RBM channels and a signal from another of the APRM channels supplies the reference signal to the second RBM channel.
The function of the RBM, to limit control rod withdrawal, is not a Safety-Related function. However, the RBM receives input signals from LPRM detectors. The cables for LPRM detectors are in scope for License Renewal and managed by GALL Program XI.E2.
Traversing In-core Probe Subsystem (TIP)
BFNs Safe Shutdown Analysis states that the safety-related functions of the TIP Subsystem are:
- 1. Provide primary containment isolation and integrity. The Safe Shutdown Analysis also states that an Active isolation function is not required.
- 2. Provide reactor coolant pressure boundary which is a passive function only.
The above functions are consistent with the intended functions provided in Section 2.3.3.33, Traversing In-Core Probe System (094), of the LRA. The TIP subsystem performs its safety-related functions without active components.
E-5 Part A Conclusion Based on the above discussions, IRM, SRM, RBM, and TIP circuits perform no intended functions as defined in 10 CFR 54.4, and therefore are not subject to an AMR.
However, based on discussions with the NRC Staff, BFN will include IRM instrumentation circuit cables within the scope of License Renewal and manage them along with LPRM cables using GALL Program XI.E2.
All other accessible neutron monitoring subsystem cables and connections will be managed by GALL XI.E1 program.
In LRA Section B.2.1.2 on Page B-18 replace the "Program Description" and "NUREG-1801 Consistency" with the following:
Program Description The Electrical Cables Not Subject To 10 CFR 50.49 Environmental Qualification Requirements Used In Instrumentation Circuits Program is a program that will manage the aging effects of sensitive, low-level signal circuits exposed to adverse localized environments caused by heat or radiation. Aging effects of Intermediate Range Monitor (IRM) and Local Power Range Monitor (LPRM) cable systems are managed by this program1.
The portion of this program that manages IRM cable system aging will be new to BFN, whereas the portion that manages LPRM cable system aging currently exists.
Technical Specification Requirements (3.3.1.1.7) impose calibration requirements on the Local Power Range Monitor circuits. The results of the calibrations will be reviewed to identify the potential existence of cable degradation. When a Local Power Range Monitor circuit is found to be out of calibration, corrective actions will be implemented. Calibrations will continue through the period of extended operation at the required frequency as specified in the BFN Technical Specifications.
NUREG-1801 Consistency The portion of the Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits Program that manages IRM cable system aging will be a new program at BFN and will be consistent with the program described in NUREG-1801 Section XI.E2.
The portion of the Electrical Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits Program that manages LPRM cable system aging will credit existing Technical Specifications requirements and, with the following exceptions, will meet the intent of the program described in NUREG-1801 Section XI.E2.
E-6 Response to Part B The Standby Gas Treatment System and the Off-Gas Treatment System are completely different systems and totally independent of each other. The Standby Gas Treatment Building and Off-Gas Treatment Building are totally different buildings which do not share power distribution systems or equipment. Standby gas treatment blowers, which are in scope for License Renewal, are not powered from Off-Gas Treatment Building Transformers A and B. They are powered as follows:
- Blower A (0-MTR-65-18) is powered from 480V Diesel Aux Bd A which is powered from 4160-480V Diesel Aux Bd Transformer TDA which is fed from 4160V Shutdown Bd A. The power for this equipment is not fed from Off-Gas Treatment Building transformers.
- Blower B (0-MTR-65-40) is powered from 480V Diesel Aux Bd B which is powered from 4160-480V Diesel Aux Bd Transformer TDB which is fed from 4160V Shutdown Bd D. The power for this equipment is not fed from Off-Gas Treatment Building transformers.
- Blower C (0-MTR-65-69) is powered from 480V Standby Gas Treatment Bd which is powered from 4160-480V Standby Gas Treatment Transformer TSG1A which is fed from 4KV Shutdown Bd 3ED. The power for this equipment is not fed from Off-Gas Treatment Building transformers.
Follow up to NRC RAI 4.7.8 Emergency Equipment Cooling Water (EECW) Weld Flaw Evaluation The applicant states that review of the EECW system indicates that continuous operation is intended; however, some interruptions have been required for maintenance and other considerations. Based on the design function of the EECW system, when and at what frequency would the system be shut down? Based on the design function and the total past history, will the number of cycles in the fatigue evaluation bound the number of cycles projected for the period of extended operation?
Please describe events, and the frequency that they have occurred, that have resulted in system operational interruptions. Should the EECW system experience more cycles than is bounded by the applicants analysis, are there procedures in place to identify this condition?
E-7 TVA Response to follow up to NRC RAI 4.7.8 The EECW system is intended to be in a continuous standby condition (i.e. under pressure-minimum flow) in both shutdown and operating plant modes. As currently designed, sections of this system may be isolated and depressurized for routine maintenance or repair. Based on operating history and future (anticipated operations) a total of 125 full pressure cycles (0 psig to design operating pressure) was selected as a conservative measure to ensure the number of fatigue cycles would not be exceeded. The preventative maintenance work orders scheduled on this system are of a periodicity of no less than 96 weeks (almost 2 years) and unless unexpected repairs are required, the system would not need to be depressurized. Using a conservatism of a little over 2 times in a year makes sense for it would be very unlikely for the same section of the EECW system to be shutdown > 2 times in a year. Please review preventative maintenance scheduled items on following page. An administrative tracking system will be developed and used to ensure that the 125 fatigue cycles will not be exceeded.
EECW PREVENTATIVE MAINTENANCE ACTIVITIES E-8 COMPONENT UNID REASON OUT OF SERVICE FREQUENCY U1 Diesel Generator Cooler A 0-HEX-082-000A1, -000A2 EDDY CURRENT TEST THE COOLERS 96 Weeks U1 Diesel Generator Cooler B 0-HEX-082-000B1, -000B2 EDDY CURRENT TEST THE COOLERS 96 Weeks U1 Diesel Generator Cooler C 0-HEX-082-000C1, -000C2 EDDY CURRENT TEST THE COOLERS 96 Weeks U1 Diesel Generator Cooler D 0-HEX-082-000D1, -000D2 EDDY CURRENT TEST THE COOLERS 96 Weeks U3 Diesel Generator Cooler A 3-HEX-082-000A1, -000A2 EDDY CURRENT TEST THE COOLERS 96 Weeks U3 Diesel Generator Cooler B 3-HEX-082-000B1, -000B2 EDDY CURRENT TEST THE COOLERS 96 Weeks U3 Diesel Generator Cooler C 3-HEX-082-000C1, -000C2 EDDY CURRENT TEST THE COOLERS 96 Weeks U3 Diesel Generator Cooler D 3-HEX-082-000D1, -000D2 EDDY CURRENT TEST THE COOLERS 96 Weeks CORE SPRAY ROOM COOLER 2A 2-CLR-064-0072 FLUSH RB CORE SPARY ROOM COOLERS A & C WITH TRI-SODIUM PHOSPHATE, CLEAN THE FAN BLADES, AND CLEAN AIR SIDE OF COOLER.
192 Weeks RHR ROOM COOLER 2C 2-CLR-064-0070 FLUSH EECW RHR ROOM COOLERS A & C WITH TRI-SODIUM PHOSPHATE, CLEAN THE FAN BLADES, AND CLEAN AIR SIDE OF COOLER.
192 Weeks RHR ROOM COOLER 2A 2-CLR-064-0068 FLUSH EECW RHR ROOM COOLERS A & C WITH TRI-SODIUM PHOSPHATE, CLEAN THE FAN BLADES, AND CLEAN AIR SIDE OF COOLER.
192 Weeks CORE SPRAY ROOM COOLER 2B 2-CLR-064-0073 FLUSH RB CORE SPARY ROOM COOLERS B & D WITH TRI-SODIUM PHOSPHATE, CLEAN THE FAN BLADES, AND CLEAN AIR SIDE OF COOLER.
192 Weeks RHR ROOM COOLER 2B 2-CLR-064-0069 FLUSH EECW RHR ROOM COOLERS B & D WITH TRI-SODIUM PHOSPHATE, CLEAN THE FAN BLADES, AND CLEAN AIR SIDE OF COOLER.
192 Weeks RHR ROOM COOLER 2D 2-CLR-064-0071 FLUSH EECW RHR ROOM COOLERS B & D WITH TRI-SODIUM PHOSPHATE, CLEAN THE FAN BLADES, AND CLEAN AIR SIDE OF COOLER.
192 Weeks RHR PUMP SEAL HEX 2A 2-HEX-074-0005 CLEAN, INSPECT AND PERFORM A PRESSURE TEST ON RHR SEAL HEAT EXCHANGER.
96 Weeks RHR PUMP SEAL HEX 2C 2-HEX-074-0016 CLEAN, INSPECT AND PERFORM A PRESSURE TEST ON RHR SEAL HEAT EXCHANGER.
96 Weeks The above PM's are driven by WO's and are the only times the related components are normally out of service (depressurized).