ML052140128
| ML052140128 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  (DPR-033) |
| Issue date: | 09/29/2006 |
| From: | Chernoff M NRC/NRR/ADRO/DORL/LPLII-2 |
| To: | Singer K Tennessee Valley Authority |
| CHERNOff, M H, NRR/DLPM, 301-415-4018 | |
| Shared Package | |
| ML062830239 | List: |
| References | |
| TAC MC4659 | |
| Download: ML052140128 (13) | |
Text
September 29, 2006 Mr. Karl W. Singer Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
SUBJECT:
BROWN FERRY NUCLEAR PLANT, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING EXCESS FLOW CHECK VALVE SURVEILLANCE TESTING INTERVALS (TAC NO. MC4659) (TS-438)
Dear Mr. Singer:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 264 to Renewed Facility Operating License No. DPR-33 for the Browns Ferry Nuclear Plant, Unit 1. This amendment is in response to your application dated October 12, 2004, as supplemented by letters dated April 27, 2005, and June 27, 2005. The amendment changes the frequency requirement for Technical Specification (TS) Surveillance Requirement 3.6.1.3.8 by allowing a representative sample (approximately 20 percent) of excess flow check valves (EFCVs) to be tested every 24 months, so that each EFCV is tested once every 120 months.
The change adopts the NRCs approved TS Task Force (TSTF) Traveler TSTF-334, Revision 2, Relaxed Surveillance Frequency for Excess Flow Check Valve Testing, dated October 31, 2000, and the BWR [Boiling-Water Reactor]/4 Standard TS, NUREG-1433, Revision 3, for primary containment isolation.
A copy of the Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Margaret H. Chernoff, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-259
Enclosures:
- 1. Amendment No. 264 to DPR-33
- 2. Safety Evaluation cc w/enclosures: See next page
September 29, 2006 Mr. Karl W. Singer Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
SUBJECT:
BROWN FERRY NUCLEAR PLANT, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING EXCESS FLOW CHECK VALVE SURVEILLANCE TESTING INTERVALS (TAC NO. MC4659) (TS-438)
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 264 to Renewed Facility Operating License No. DPR-33 for the Browns Ferry Nuclear Plant, Unit 1. This amendment is in response to your application dated October 12, 2004, as supplemented by letters dated April 27, 2005, and June 27, 2005. The amendment changes the frequency requirement for Technical Specification (TS) Surveillance Requirement 3.6.1.3.8 by allowing a representative sample (approximately 20 percent) of excess flow check valves (EFCVs) to be tested every 24 months, so that each EFCV is tested once every 120 months.
The change adopts the NRCs approved TS Task Force (TSTF) Traveler TSTF-334, Revision 2, Relaxed Surveillance Frequency for Excess Flow Check Valve Testing, dated October 31, 2000, and the BWR [Boiling-Water Reactor]/4 Standard TS, NUREG-1433, Revision 3, for primary containment isolation.
A copy of the Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Margaret H. Chernoff, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-259
Enclosures:
- 1. Amendment No. 264 to DPR-33
- 2. Safety Evaluation cc w/enclosures: See next page Distribution:
PUBLIC LPL2-2 RF RidsNrrPMMChernoff RidsNrrPMEBrown RidsNrrDorlLpl2-2 RidsOgcRp RidsNrrLABClayton (Hard Copy)
RidsOgcRp RidsAcrsAcnwMailCenter CDoutt GHill (2 copies)
RidsNrrDorlDpr RidsRgn2MailCenter RidsNrrDraApla RidsNrrDirsItsb Package: ML062830239 TS: ML062840069 ADAMS ACCESSION NO. ML052140128
- No Legal Objection NRR-058 OFFICE LPL2-2/PM LPL2-2/LA APLA/BC OGC LPL2-2/BC NAME MChernoff BClayton MRubin by memo MZobler*
LRaghavan DATE 8/15/06 8/15/06 7/13/05 8/15/06 9/29/06 OFFICIAL RECORD COPY
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 264 Renewed License No. DPR-33 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated October 12, 2004, as supplemented April 27 and June 27, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (I) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-33 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 264, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
L. Raghavan, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 29, 2006
ATTACHMENT TO LICENSE AMENDMENT NO. 264 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Replace Page 3 of Renewed Operating License DPR-33 with the attached Page 3.
Revise the Appendix A Technical Specifications by removing the page identified below and inserting the attached page. The revised page is identified by the captioned amendment number and contains a marginal line indicating the area of change.
REMOVE INSERT 3.6-16 3.6-16
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 264 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-259
1.0 INTRODUCTION
By letter October 12, 2004, (Agency Document Management System (ADAMS) Accession No. ML042870100) as supplemented by letters dated April 27, 2005 (ADAMS Accession No. ML051300462), and June 27, 2005 (ADAMS Accession No. ML051790015), the Tennessee Valley Authority (TVA, the licensee) submitted a request for changes to the Browns Ferry Nuclear Plant (BFN), Unit 1, Technical Specifications (TSs). The requested changes would revise TS Surveillance Requirement (SR) 3.6.1.3.8 to relax the 24-month excess flow check valve (EFCV) surveillance frequency by limiting the number of tests to a representative sample every 24 months such that each EFCV will be tested at least once every 120 months.
The proposed change adopts the U.S. Nuclear Regulatory Commission (NRC) approved TSs, Task Force (TSTF) Traveler TSTF-334, Revision 2, Relaxed Surveillance Frequency for Excess Flow Check Valve Testing, dated October 31, 2000, and the BWR [Boiling Water Reactor]/4 Standard TSs, NUREG-1433, Revision 3, for primary containment isolation valves.
The April 27 and June 27, 2005, letters provided clarifying information that did not expand the scope of the original application or change the initial proposed no significant hazards consideration determination.
2.0 REGULATORY EVALUATION
The basis for the request is the high degree of reliability shown by the EFCVs and the low consequences of an EFCV failure. The supporting analysis for the licensees conclusion is based on General Electric Nuclear Energy (GENE) Topical Report NEDO-32977-A, Excess Flow Check Valve Testing Relaxation, dated June 2000. The topical report provided:
(1) estimate of steam release frequency into the reactor building due to a break in an instrument line concurrent with an EFCV failure to close, and (2) assessment of the radiological consequences of such a release. The BWR Owners Group (BWROG) concluded that EFCVs testing intervals could be extended up to 10 years based on the topical report reliability and consequence analysis without significantly affecting plant risk. The BWROG suggested a staggered test interval based on actual valve performance with each valve being tested at least once every 10 years. The NRC staff accepted the generic applicability of the topical report as indicated in its Safety Evaluation (SE) dated March 14, 2000, and agreed that the EFCV test interval could be extended to as much as 10 years. The NRC staff also noted that licensees adopting the topical report must have a failure feedback mechanism and corrective action program to ensure that EFCV performance continues to be bounded by the topical report results. Additionally, each licensee is required to perform a plant-specific radiological dose assessment and EFCV failure rate and release frequency analysis to confirm that their facility is bounded by the generic analysis of the topical report. TSTF-334, Revision 2, is applicable only for those plants bounded by the analyses presented by NEDO-32977-A, and are subject to EFCV performance and corrective action criteria to be developed by the licensee.
The NRC staff reviewed the licensees submittal for conformance to the March 14, 2000, staff SE to Topical Report NEDO-32977-A and the guidance of approved TSTF-334 Revision 2. The NRCs evaluation concerned itself with the following areas: (1) EFCV failure rate and release frequency, (2) the licensees failure feedback mechanism and corrective action program, (3) radiological dose assessment, and (4) conformance of the revised TS to generic TS guidance.
2.1 Applicable Regulations Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36 requires that all operating licenses for nuclear reactors must include the TSs for the subject plant.
The limiting condition for operation, along with the required completion times (CTs) are specified for each system included in the TSs.
Maintenance Rule, 10 CFR 50.65(a)(4), as it relates to the proposed EFCV CT configuration, requires the monitoring of the performance or condition of structures, systems and components against licensee-established goals.
Appendix A of 10 CFR Part 50 - General Design Criteria (GDC) 55, Reactor Coolant Pressure Boundary Penetrating Containment, requires that each line that is part of the reactor coolant pressure boundary and that penetrates primary containment shall be provided with containment isolation valves.
Appendix A of 10 CFR Part 50 - GDC 56, Primary Containment Isolation, requires that each line connecting directly to containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as listed, unless it can be demonstrated that containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis.
2.2 Applicable Regulatory Criteria/Guidelines Regulatory Guide (RG) 1.11, Instrument Lines Penetrating Primary Reactor Containment, provides for the use of EFCVs to satisfy the requirements of Appendix A of 10 CFR Part 50 - GDC 55 and 56 for automatic isolation capability of lines penetrating containment while maintaining a highly reliable capability to monitor important parameters inside containment.
3.0 TECHNICAL EVALUATION
EFCVs are installed in BWR instrument lines penetrating the primary containment boundary to limit the release of fluid in the event of an instrument line break. RG 1.11, Instrument Lines Penetrating Primary Reactor Containment, provides guidance on the implementation of GDC 55 and 56 for instrumentation lines that penetrate primary reactor containment and are part of the reactor coolant pressure boundary. As stated by RG 1.11, EFCVs, in combination with flow restricting features (line size or orifice), satisfy the requirements of GDC 55 and 56 for automatic isolation capability, maintain the reliability of the connected instrumentation, and ensure the functional performance of secondary containment in the event of an instrumentation line rupture. Examples of EFCV installations include reactor pressure vessel level and pressure instrumentation, main steam line flow instrumentation, recirculation pump suction pressure, and reactor core isolation cooling steam line flow instrumentation. EFCVs are not required to close in response to a containment isolation signal and are not required to operate under post-loss-of-coolant accident conditions.
3.1 Proposed TS Change BFN Unit 1 TS SR 3.6.1.3.8 currently requires verification of the actuation capability of each reactor instrumentation line EFCV every 24 months. The SR demonstrates that each reactor instrumentation line EFCV is operable by verifying that the valve actuates to restrict flow to within limits. The proposed change revises TS SR 3.6.1.3.8 to relax the 24-month EFCV surveillance frequency by limiting the number of tests to a representative sample (approximately 20 percent) every 24 months, such that each EFCV will be tested at least once every 120 months. The representative sample consists of approximately equal number of EFCVs being tested every 24 months, such that each EFCV is tested at least once every 120 months.
The basis for the request is the high degree of reliability shown by the EFCVs and the low consequences of an EFCV failure. The supporting analysis for the licensees conclusion is based on GENE Topical Report NEDO-32977-A, Excess Flow Check Valve Testing Relaxation, dated June 2000. The topical report provided: (1) estimate of steam release frequency into the reactor building due to a break in an instrument line concurrent with an EFCV failure to close, and (2) assessment of the radiological consequences of such a release. The BWROG concluded that EFCVs testing intervals could be extended up to 10 years based on the topical report reliability and consequence analysis without significantly affecting plant risk.
The BWROG suggested a staggered test interval based on actual valve performance with each valve being tested at least once every 120 months. The NRC accepted the generic applicability of the topical report via SE dated March 14, 2000, and agreed that the EFCV test interval could be extended to as much as 120 months. The NRC staff also noted that licensees adopting the topical report must have a failure feedback mechanism and corrective action program to ensure that EFCV performance continues to be bounded by the topical report results. Additionally, each licensee is required to perform a plant-specific radiological dose assessment and EFCV failure rate and release frequency analysis to confirm that their facility is bounded by the generic analysis of the topical report.
3.2 EFCV Failure Rate and Release Frequency In the topical report, EFCV reliability was evaluated based on testing experience provided by 12 different BWR plants. The composite data indicated that EFCVs are very reliable. The data represented 12,424.5 valve-years of operation with a total of 11 failures noted. The EFCV composite failure rate was 1.67E-07/hour and was referenced as the upper limit failure rate in the topical report.
The NRC staff noted in its review of the report that the BWROG assumed the EFCV failure rate was constant over time and did not account for potential age-related degradation in the EFCV failure rate. Additionally, the NRC staff questioned the use of an instrument line break frequency based on WASH-1400 and not on more current data. To address this concern, the BWROG request for additional information (RAI) response included an updated instrument line failure frequency of 3.52E-05 failures/year based on the Electric Power Research Institutes Technical Report No. 100380, Pipe Failures in U.S. Commercial Nuclear Power Plants, dated July 1992. This value is 6.6 times greater than the value calculated in the topical report using WASH-1400 data. The BWROG RAI response also assumed the observed EFCV failures were five times the actual observed number (55 vs. 11) listed in the topical report. The additional impact of an increase in instrument line failure frequency and a fivefold increase in EFCV failures assumed by the BWROG RAI response demonstrated that release frequencies remained low with limited impact.
To estimate the release frequency initiated by an instrument line break, two factors are considered: (1) the instrument line break frequency downstream of the EFCV, and (2) the probability of the EFCV failing to close. The licensee was a participating utility in the development of the EFCV topical report. The BFN Units 2 and 3 data were found to be consistent in both the time sampled and EFCV reliability when compared to the topical report data, once additional data and trends were considered. The BFN Unit 1 plant-specific EFCV failure and release rates are not available because of the extended shutdown of BFN Unit 1.
To extend the EFCV surveillance intervals for BFN Unit 1 the licensee has used the previous data from the BFN Units 2 and 3 EFCV review and additional data available from more recent BFN Units 2 and 3 refueling outages. The basis for using BFN Units 2 and 3 data is the fact that all three BFN units use only one type of EFCV for instrument lines connected to the reactor coolant pressure boundary. In addition, the stated EFCV operating conditions and environment are similar across valve installations and BFN units.
In the NRCs previous review of the EFCV surveillance extension for BFN Units 2 and 3 the NRC staff noted that both BFN Units 2 and 3 experienced EFCV failures during plant restart.
Specifically, there were 21 EFCV failures during BFN Unit 2 restart and 5 EFCV failures during BFN Unit 3 restart. In the original EFCV SE for BFN Units 2 and 3, the NRC staff was concerned that EFCV failures may not be bounded by the topical report results. The licensee stated that the failures noted upon restart were the result of 6-and 10-year extended outages respectively. The licensee considered these failures to be a direct result of instrument line conditions during extended shutdown, including crud build-up, test methods, and oxidation of valve internals causing the valves to stick during testing. Conditions during operation are significantly different from those during shutdown that caused the valve failures.
Subsequent to these failures the licensee implemented improvements to procedures for EFCV testing and EFCV test methodology. Additional job training was performed with task qualifications documented and continuity of plant personnel involved with the test being maintained. EFCV testing is handled as a complex, infrequently performed, test or evolution to ensure an appropriate level of management oversight. For the BFN Unit 1 EFCV amendment request, the licensee provided additional plant operational data that covered the time period subsequent to the BFN Units 2 and 3 EFCV surveillance extension request. The additional data indicates that the EFCVs remain reliable with no additional failures noted since the three EFCV failures experienced after the startup of BFN Units 2 and 3 (two failures for Unit 2 and one failure for Unit 3). Therefore, continued EFCV testing performed by the licensee has shown a significant reduction in EFCV failures since initial plant restart.
To prevent the high initial EFCV failures experienced on BFN Unit 2 from occurring on BFN Unit 1, the licensee will flush instrument sensing lines, and inspect, clean, and test all EFCVs prior to BFN Unit 1 startup, replacing failed EFCVs as required. Testing will be performed during plant hydrostatic testing prior to restart. The licensee also stated that the test methodology was improved to include followup bench testing following an EFCV failure.
The BFN EFCV failure data were updated to reflect additional testing performed since the topical report and the EFCV staff SE for BFN Units 2 and 3 were issued. The BFN Unit 1 submittal indicates EFCV failure rates that are somewhat higher than presented in the topical report but improved from the previous BFN Units 2 and 3 evaluation. These higher frequencies are directly related to reduced operating time with additional failures compared to the topical report composite values. The failures noted occurred either upon BFN Units 2 and 3 startup and are attributable to line conditions during shutdown or in the refueling outages immediately following BFN Units 2 and 3 restart. Since then, testing has shown a significant reduction in EFCV failures with no additional EFCV failures noted by the licensee through Refueling Cycle 12 on BFN Unit 2 and Cycle 11 on BFN Unit 3. The performance of EFCVs in later cycles is representative of the failure rates noted in the topical report and is within the acceptance criteria established for EFCV performance. Based on the above, it is expected that the performance of BFN Unit 1 EFCVs should be similar to the performance shown for BFN Units 2 and 3 after restart. This conclusion is based on the fact that the BFN Unit 1 EFCVs are of the same manufacture, see similar environmental and operating conditions, and the licensee has established appropriate test plans, procedures, performance monitoring, and corrective action programs to ensure EFCV performance remains with the established acceptance criteria.
As demonstrated in GENE Topical Report NEDO-32977-A, the impact of an increase in the EFCV surveillance test interval to 10 years results in an instrument line release frequency considered by the NRC staff to be sufficiently low, especially since the consequences of an EFCV failure are bounded by previous licensee analysis and, therefore, are highly unlikely to lead to core damage. Additionally, the licensees evaluation results (including the plant-specific EFCV failure data and release frequency) is consistent with the topical report results. The NRC staff concludes that the release frequency associated with the BFN Unit 1 request for relaxation of EFCV surveillance testing will be sufficiently low and, therefore, acceptable.
3.3 Licensees Failure Feedback Mechanism and Corrective Action Program The topical report established that each plants corrective action program must evaluate equipment failures and establish appropriate corrective actions. These programs ensure that meaningful feedback data is acquired so that appropriate corrective action may be taken with regard to EFCV performance. The NRC staff noted that the topical report does not provide a specific failure feedback mechanism, but does state that a plants corrective action program must evaluate equipment failures and establish appropriate corrective actions. The BWROG responded to the NRC staff RAI question concerning failure feedback by stating that each licensee who adopts the relaxed surveillance intervals recommended by the topical report should ensure that an appropriate feedback mechanism responsive to EFCV failure trends is in place.
The licensee provided input on EFCV performance criteria and the EFCV corrective action program confirming that any EFCV failure will be evaluated under the BFN corrective action program. The licensee previously stated that the Maintenance Rule program was revised to provide a means of monitoring EFCV reliability. The licensee also established EFCV performance criterion in accordance with the requirements of the licensees Maintenance Rule program. The acceptance criterion was established as a number of test failures over a specified time interval (two failures per 2-year rolling period). The acceptance criterion is contained in the BFN Maintenance Rule program technical instruction 0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -10 CFR 50.65. In addition, TSTF-334, Revision 2 states that any EFCV failure will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability is maintained.
The licensee has developed this program to ensure that EFCV performance will remain consistent with the extended surveillance interval assumptions and adverse trends in EFCV performance are identified and corrected. The NRC staff considers the licensees program to account for potential changes in EFCV failure rates to be acceptable and satisfies TSTF-334 performance and corrective action criteria.
The topical report established that each plants corrective action program must evaluate equipment failures and establish appropriate corrective actions. These programs ensure that meaningful feedback data is acquired so that appropriate corrective action may be taken with regard to EFCV performance. The licensee provided information regarding EFCV performance criteria and the EFCV corrective action program. The NRC staff finds the licensees program to be in conformance with TSTF-334, Revision 2, and the topical report and, thus, is also acceptable to the NRC staff.
3.4 Radiological Dose Assessment The licensee performed a plant-specific radiological consequence dose assessment for the off-site and control room for an instrumentation line break outside of the drywell. A manual reactor scram is assumed to occur 10 minutes after the instrument line break. The licensee further assumed that the reactor coolant iodine activity was at equilibrium TS limit of 3.2 micro curies per gram of dose equivalent iodine-131 and that an iodine spike of 500 times the equivalent iodine release rate to the reactor primary coolant occurred. The analysis does not credit the EFCVs for isolating the break but does assume the discharge of reactor water is through an instrument line with a 1/4-inch flow-restricting orifice for the duration of the event.
The postulated coolant leakage is within the capability of the reactor coolant makeup systems should an EFCV fail to close. In addition, based on the data presented by the licensee and the topical report, the BFN EFCV release frequency increase is not significant in that this frequency is lower than the large break loss-of-coolant accident, which has the potential to cause core damage but the failure of an EFCV failing to close concurrent with an instrument line break does not.
The topical report stated that the magnitude of release through an instrument line would be within the pressure control capacity of reactor building ventilation systems and that the integrity and functional performance of secondary containment and standby gas treatment system following an instrument line break would continue to be met. The licensee confirmed in their submittal that should an EFCV fail, the restricting orifice or line restriction limits the steam release and the integrity and functional performance of secondary containment will be maintained.
The resulting radiological consequences for offsite and control room are less than small fractions of the dose criteria specified in 10 CFR 50.67, Accident Source Term. As a result, a failure of an EFCV to close is bounded by the licensees previous analysis. The radiation dose consequences for an instrument line break are, therefore, not impacted by the proposed change. Therefore, the NRC staff agrees that the current licensing basis remains applicable for the proposed EFCV surveillance interval, with regard to the potential radiological consequences of an instrument line break with failure of the EFCV to isolate or result in a significant increase in the consequences of an accident previously evaluated.
Based on the above, the NRC staff finds the proposed change to relax the BFN Unit 1 instrument line EFCV surveillance frequency by allowing a representative sample of EFCVs to be tested every 24 months, with all EFCVs being tested at least once every 120 months to be consistent with TSTF-334 generic guidance, Topical Report NEDO-32977-A, the NRCs March 14, 2000, SE, and is, therefore, acceptable
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Alabama State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (70 FR 15948). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Clifford K. Doutt Date: September 29, 2006