ML042870100
| ML042870100 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/12/2004 |
| From: | Abney T Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TVA-BFN-TS-438 | |
| Download: ML042870100 (22) | |
Text
October 12, 2004 TVA-BFN-TS-438 10 CFR 50.90 U.S. Nuclear Regulatory Commission Mail Stop: OWFN P1-35 ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of )
Docket No. 50-259 Tennessee Valley Authority )
BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 1 - TECHNICAL SPECIFICATIONS (TS) CHANGE 438 - REVISION TO EXCESS FLOW CHECK VALVE (EFCV) SURVEILLANCE TESTING FREQUENCY
Reference:
TVA letter, T.E. Abney to NRC, Browns Ferry Nuclear Plant (BFN) Unit 1 - Technical Specification Change (TS) 433 - 24 Month Fuel Cycle.
Pursuant to 10 CFR 50.90, TVA is submitting a request for a TS change (TS 438) to license DPR-33 for BFN Unit 1. The proposed TS change will revise the frequency of Surveillance Requirement 3.6.1.3.8 by testing a representative sample (approximately 20 percent) of EFCV every 24 months, so each EFCV is tested once every 120 months.
The proposed TS change is necessary to support the restart of Unit 1. This request is consistent with the BWR/4 Standard Technical Specifications, NUREG-1433, Revision 3, for Primary Containment isolation valves and the Browns Ferry Units 2 and 3 TS. The change is also necessary to ensure consistency with the current Units 2 and 3 TS.
U.S. Nuclear Regulatory Commission Page 2 October 12, 2004 TVA has determined no significant hazards considerations are associated with the proposed change and the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health.
TVA has previously requested approval to revise the Unit 1 TS (TS 433 - Reference) to decreases the frequency of once-per-cycle from 18 months to 24 months in the affected TS Surveillance Requirements. This proposed change assumes TS 433 has been approved. The review and approval of this TS change should be staged and coordinated by the NRC in such a manner as not to invalidate this assumption.
TVA requests approval of this TS change by September 30, 2005 and that the implementation of the revised TS be within 60 days of NRC approval.
provides TVAs evaluation of the proposed change. provides a mark-up of the proposed TS changes. provides the revised TS pages with the proposed changes incorporated.
There are no regulatory commitments associated with this submittal. If you have any questions about this change, please contact me at (256)729-2636.
I declare under penalty of perjury that the foregoing is true and correct. Executed on October 12, 2004.
Sincerely, Original signed by:
T. E. Abney Manager of Licensing and Industry Affairs
Enclosures:
- 1. TVA Evaluation of Proposed Change
- 2. Proposed Technical Specification Changes (mark-up)
- 3. Proposed Technical Specification Changes (retyped)
U.S. Nuclear Regulatory Commission Page 4 October 12, 2004 cc: (Enclosures)
State Health Officer Alabama State Department of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-3415 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, AL 35611-6970 Kahtan N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739
U.S. Nuclear Regulatory Commission Page 4 October 12, 2004 SMK:BAB Enclosures cc (w/o Enclosures):
A. S. Bhatnagar, LP 6A-C J. C. Fornicola, LP 6A-C D. F. Helms, BR 4T-C R. F. Marks, PAB 1C-BFN R. G. Jones, NAB 1A-BFN K. L. Krueger, POB 2C-BFN J. R. Rupert, NAB 1A-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 (with Enclosures) s:lic/submit/TechSpec/TS 438 excess flow.doc
E1-i Browns Ferry Nuclear Plant (BFN)
Technical Specification (TS) Change TS 438 Revision to Excess Flow Check Valve (EFCV)
Surveillance Testing Frequency TVA Evaluation of Proposed Change INDEX SECTION TOPIC PAGE 1.0 Description...................................
1 2.0 Proposed Amendment............................ 1 3.0 Background.................................... 1 3.1 Reason for the Proposed Unit 1 Amendment...... 2 3.2 Description of Excess Flow Check Valve......... 3 3.3 Comparison with Previous Technical Specification changes for Unit 2 and 3......................
4 4.0 Technical Analysis............................ 5 5.0 Regulatory Safety Analysis.................... 10 5.1 No Significant Hazards Consideration.......... 10 5.2 Applicable Regulatory Requirements/Criteria... 12 6.0 Environmental Consideration................... 12 7.0 References.................................... 13
E1-1
1.0 DESCRIPTION
This letter requests an amendment to Operating License DPR-33 for BFN Unit 1. The current TS requires each reactor instrumentation line EFCV be actuated to the isolation position every 24 months.
The proposed amendment will revise the frequency of Surveillance Requirement (SR) 3.6.1.3.8 by testing a representative sample (approximately 20 percent) of EFCV every 24 months, so each EFCV is tested once every 120 months.
This request is consistent with the BWR/4 Standard Technical Specifications, NUREG-1433, Revision 3, for Primary Containment Isolation Valves and the Browns Ferry Units 2 and 3 TS.
The proposed TS change is necessary to support the restart of Unit 1. The change is also necessary to ensure consistency with the current Units 2 and 3 TS. TVA requests the amendment be approved by September 30, 2005.
2.0 PROPOSED AMENDMENT The proposed amendment will revise the frequency of SR 3.6.1.3.8 on Page 3.6-16 by allowing a representative sample (approximately 20 percent) of EFCV to be tested every 24 months, so each EFCV is tested once every 120 months. The deleted text is shown with strikethrough, and changed or added text is shown in bold italics. The revised SR will read as follows:
SURVEILLANCE FREQUENCY SR 3.6.1.3.8 Verify each a representative sample of reactor instrumentation line EFCV actuates to the isolation position on a simulated instrument line break signal.
24 months provides a mark-up of the proposed TS changes. provides the revised TS pages with the proposed changes incorporated.
3.0 BACKGROUND
The proposed change implements Technical Specifications Task Force (TSTF) Traveler TSTF 334, Revision 2, "Relaxed Surveillance Frequency for Excess Flow Check Valve Testing." The proposed
E1-2 change was internally approved by NRC staff in Reference 1. The Safety Evaluation for the associated General Electric Topical Report was issued in Reference 2. Applicability was limited to those facilities encompassed by the analyses performed in support of the topical report, and are subject to performance and corrective action criteria to be developed by the licensee.
These requirements are addressed in this submittal.
TVA has previously implemented similar changes to the BFN Units 2 and 3 TS. On August 11, 2000, TVA requested changes to Units 2 and 3 TS SR 3.6.1.3.8 by allowing a representative sample (approximately 20 percent) of EFCV to be tested every 24 months, so each EFCV is tested once every 120 months (Reference 3). On October 5, 2000, the NRC requested additional information, which TVA provided on October 20, 2000 (References 4 and 5, respectively). NRC approved the amendments for Units 2 and 3 on January 29, 2001 (Reference 6).
The overall content of this submittal is based on TVAs original application for Units 2 and 3 and the response to NRCs request for additional information. Included at the end of this section is a comparison of the proposed change, reason for change, background information, and technical analysis submitted in support of this proposed amendment with the information provided by TVA and approved by NRC for the Units 2 and 3 license amendments. This section also contains a discussion of:
- 1.
The reasons for the proposed Unit 1 amendment;
- 2.
A description of EFCV; and
- 3.
TVAs risk assessment for the proposed change.
3.1 Reason for the Proposed Unit 1 Amendment TVA is requesting an amendment to the Unit 1 TS to reduce the number of EFCVs tested as part of refueling outage activities.
Instrument line EFCVs which connect to the reactor coolant pressure boundary are normally tested during the reactor pressure vessel system leakage test, which is performed as part of refueling outage activities. The EFCVs are designed to close on a failure of an instrument line downstream of the valve. Testing of the EFCVs typically requires that the reactor be pressurized to normal operating pressure, opening an instrument drain valve and observing valve closure. Testing has historically required approximately 140 man-hours to complete and can be a critical path item during refueling outages. This places an undue burden on the TVA staff without a commensurate increase in plant safety.
E1-3 Reducing the number of valves tested, as with any reduction in maintenance, inherently reduces the risk of industrial and occupational hazards, including inadvertent exposure to radioactive fluids. Furthermore, there is a consequential reduction in radioactive waste generated during testing activities.
3.2 Description of Excess Flow Check Valve BFN utilizes only one type of EFCV (Marotta FVL 16D) in instrument lines connected to the reactor coolant pressure boundary. The operating environment and conditions experienced by any one valve are similar for all valves in the population.
The Marotta FVL 16D is a flow limiting valve designed to shutoff when downstream flow increases to a predetermined rate. The valve has a controlled leakage rate when in the shutoff position.
This allows the poppet to return to the open position when flow through the valve is reduced to zero. These valves are self contained, not adjustable, and no maintenance is required under normal conditions.
Instrumentation piping connected to the reactor primary system which leaves the primary containment is dead ended at instrument racks located in the Reactor Building. These instrument lines are provided with a manual block valve and an EFCV both of which are located outside primary containment. There are 67 EFCVs installed on BFN Unit 1. Except for the jet pump sensing lines, a one fourth (1/4)-inch orifice has been installed in instrument lines that penetrate the primary containment boundary into the secondary containment (Reactor Building area). Sense lines for jet pump flow within the reactor vessel, including the reactor vessel penetration, are constructed of one fourth (1/4)-inch sensing pipe. The one fourth (1/4)-inch sense line, effectively provides the same flow area as the one fourth (1/4)-inch orifice in the other instrument lines. This design limits the release of reactor coolant in the event of an instrument line break outside primary containment.
3.3 Risk Assessment of the Proposed Change The release frequency is the product of the instrument line break frequency and EFCV failure rates. An instrument line break frequency of 3.52E-05 per year has been calculated for BFN. The EFCV failure rate was derived based on the assumption that there will be a five fold increase in failures at BFN. Failure rate for the ten year interval is obtained using a chi-square distribution considering 5 times the normal failure rate in the ten year period. TVA found for a two year surveillance interval (24 month fuel cycle) the release frequency would be approximately 4.93E-05 events per year and for a ten year
E1-4 surveillance interval the release frequency would be 2.47E-04 events per year.
The specific values for the two and ten year surveillance interval are sufficiently low that it can be concluded a change in surveillance test frequency has minimal impact on EFCV reliability.
The release frequency, considering a ten year surveillance frequency BFN, is equivalent to approximately one event every 4,000 years.
3.4 Comparison with Previous Technical Specification changes for Unit 2 and 3 TVA has compared the proposed change, reason for change, background information, and technical analysis submitted in support of this proposed amendment with the information provided by TVA and approved by NRC in References 1 through 4 for the revision to the Units 2 and 3 TS to allow a representative sample (approximately 20 percent) of EFCV to be tested every 24 months, so each EFCV is tested once every 120 months. The comparison for each of these areas is provided below:
The proposed change to the Unit 1 TS is the same change as proposed and approved for Units 2 and 3.
The underlying reason for the Unit 1 TS change is the same as previously submitted for the Units 2 and 3 TS change (i.e., Testing of every EFCV has historically required approximately 140 man-hours to complete and can be a critical path item during refueling outages. This places an undue burden on the TVA staff without a commensurate increase in plant safety.). In addition, TVA needs to maximize consistency between the Unit 1 and Units 2 and 3 TS, operations and maintenance practices prior to restarting Unit 1.
The background information provided in support of the Unit 1 TS change incorporates the same elements previously submitted in support of the Units 2 and 3 TS change.
The technical analysis submitted for this Unit 1 TS change incorporates the same elements previously submitted in support of the previous TS changes for Units 2 and 3.
However, several items required updated since that time:
E1-5 The Units 2 and 3 change was based on pre-uprate conditions (Rated Thermal Power of 3293 MWt). The Unit 1 TS change are based on Extended Power Uprate conditions (3952 MWt).
The EFCV failure rate incorporates two additional cycles of test data from Units 2 and 3 surveillances.
4.0 TECHNICAL ANALYSIS
The leakage from a postulated broken instrument line outside containment is reduced by design to the maximum extent practical, consistent with instrument response requirements. The rate and extent of coolant loss is well within the capability of reactor coolant make-up systems. The integrity and functional performance of secondary containment and the Standby Gas Treatment System will be maintained. A break in the portion of an instrument line between the containment and EFCV located outside primary containment and the direct blowdown to the reactor building was considered. It was concluded in the June 26, 1972, Safety Evaluation for the Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3, the isolation provisions for instrument lines penetrating primary containment were adequately designed and meets the intent of NRC Safety Guide 11, Instrument Lines Penetrating Primary Containment Backfitting Considerations (Regulatory Guide 1.11, dated February 17, 1972).
The safety objective of the Primary Containment Isolation System is to provide the capability, in the event of the postulated loss-of-coolant accident, to limit the release of fission products to the plant environs so that offsite and control room operators doses would be within the limits of 10 CFR 50.67.
Isolation of all pipes or ducts which penetrate the primary containment is required to maintain leakage within permissible limits.
An analysis was performed to determine the feasibility of relaxing the current TS SR to test each reactor instrumentation line EFCV every 24 months. The proposed change would allow a representative sample of EFCVs to be tested each 24 months. Each EFCV would be tested at least once every 120 months (nominal).
This test program is consistent with that described in General Electric NEDO-32977-A (Boiling Water Reactor Owners Group (BWROG)
Topical Report B21-00658-01), Excess Flow Check Valve Testing Relaxation, dated November, 1998, (revised through June, 2000).
This report was approved by the staff on March 14, 2000.
E1-6 BWROG Topical Report B21-00658-01 was reviewed, along with the licensing requirements, operational experience, and consequences associated with the testing requirements for EFCVs in instrument lines connecting to the reactor coolant pressure boundary. The report concluded the change in the test frequency had insignificant impact on valve reliability. The BWROG report also concluded that the demonstrated reliability of EFCVs coupled with low consequences of EFCV failure provided adequate justification for extending the test interval up to once every 120 months.
The effect of extending the EFCV test interval is a corresponding increase in the potential for a release. However, even with a 120 month test interval, the release frequency from an individual line continues to remain very low. Also, since the EFCVs are located in secondary containment (Reactor Building), any release from a failed EFCV would be treated by the Standby Gas Treatment System providing additional mitigation of any postulated offsite release from a broken instrument line.
A. RADIOLOGICAL DOSE ASSESSMENT TVA performed a plant specific dose assessment to determine the off-site and control room dose due to an instrument line break outside of the drywell. The reactor was assumed to be operating at normal pressure and temperature. A manual reactor scram is assumed to occur ten minutes following the line break. The reactor coolant was at equilibrium TS limits for operation of 3.2 µCi/gm I-131 equivalent. As a consequence of the reactor scram and subsequent depressurization an iodine spike of 500 times the equilibrium release rate was incorporated in the model.
When evaluating the radiological impact from an instrument line break, credit was taken for a 1/4-inch flow restriction.
The postulated leak is modeled as flow from the reactor coolant system to either the sump (the non-flashed liquid fraction) or the reactor building (the flashed steam fraction). The leak flow rates and fluid pressure are taken from Figure 8-1 on Page 8-2, Total Mass Flow Rate From Instrument Line Break - 1/4 inch Orifice, and Figure 8-2 on Page 8-3, Reactor Pressure During Normal Shutdown, of General Electric NEDO-21143-1, Radiological Accident Evaluation - The CONAC03 Code, dated December 1981. To establish the flashing fraction the reactor pressure at a given time during a postulated line break is obtained from Figure 8-2. Once the flashing fraction is established, using the steam tables, the enthalpy at saturation at the reactor pressure (1055 psia to 22 psia) is then calculated.
Assuming constant enthalpy, the flashing fraction will be the quality of the reactor fluid at atmospheric pressure
E1-7 (14.4 psia for BFN). Once the mass release and fractions are established, the liquid and steam release is determined.
The computer code STP was used to determine the time dependent releases. The computer code COROD was used to determine the control room operator doses and computer code FENCDOSE was used to determine the offsite doses. These computer codes are part of the BFN licensing basis as described in Chapter 14 of the FSAR. The release rate and pressure of the water were derived from GE NEDO-21143-1, Radiological Accident Evaluation - The CONAC03 Code, dated December 1981. GE NEDO-21143-1 assumes Operator action occurs at ten minutes. The Standby Gas Treatment system is initiated followed by a plant shutdown. The results of the assessment are shown below. The doses are in REM.
Dose Assessment Resulting From EFCV Failure Control Room 2-Hour Exclusion Boundary 30-Day Low Population Zone Inhalation (ICRP-30) 2.286x10-1 5.516x10-1 4.024x10-1 Gamma 2.200x10-4 1.244x10-2 1.842x10-2 Beta 1.143x10-3 2.986x10-3 6.570x10-3 TEDE 7.532x10-3 5.366x10-2 4.378x10-2 As shown in the table above, the total control room Total Effective Dose Equivalent (TEDE) doses are less than the 5 REM limit and the offsite TEDE doses are less than 10% of the 25 REM limit (10 CFR 50.67).
B. EFCV FAILURE RATE The reliability of EFCVs was evaluated based on testing experience provided by 12 different BWR plants. The composite data indicated that EFCVs are very reliable. It can be noted that the data shown in the BWROG report documents that BFN experienced 21 EFCV failures during the Unit 2 restart effort and 5 EFCV failures during the Unit 3 restart effort. The testing followed a 6 and 10 year extended outage respectfully. TVA attributed this high failure rate to crud build up and valve sticking, test methodology, lack of experience of the test personnel, and one broken valve spring. TVA is addressing crud build up and potential sticking on Unit 1 restart by flushing sense lines and inspecting 100 percent of the EFCVs (i.e. 67 valves) and replacing as necessary. There are improvements in test methodology which include bench testing of valves following failure to determine if the valve failures were actual. On the job training (OJT) of personnel and task qualifications are documented and continuity of personnel involved with the tests has been maintained. Additionally,
E1-8 EFCV testing is handled as a complex infrequently performed test or evolution which ensures an appropriate level of management oversight.
The lessons learned from Unit 2 restart have been effective in reducing the number of EFCV failures. This is supported by actual performance data. Since initial restart of Units 2 and 3 in 1991 and 1995 respectfully, TVA has experienced only 3 EFCV failures. The three failures were unrelated and there was no common root cause. TVA reviewed the surveillance test results from the last eleven refueling outages (six outages on Unit 2 and five outages on Unit 3.
The results of this review are provided in the tables below:
BFN EFCV Failures Cycle 7 Cycle 8 Cycle 9 Cycle 10 Cycle 11 Cycle 12 Unit 2 1
1 0
0 0
0 Unit 3 1
0 0
0 0
NA Additionally, TVA calculated a plant specific conservative failure rate from the eleven refueling outages and an 18 month fuel cycle. The last 4 cycles have been 24 months in duration but 18 months duration is used in the calculation.
This is shown below:
BFN EFCV Failures Rate Operating time (years) 1 Operating Time (hours) 1 Number of Failures Estimate Failure Rate/hour 2 Upper Limit Failure Rate/Hour 2 1105.5 9.68E+06 3
3.10E-07 8.01E-07 1105.5 9.68E+06 152 1.55E-06 2.39E-06 As shown in the tables, BFN has experienced 3 EFCV failures in the previous 9.68+06 hours of reactor operation. The calculated upper limit value for BFN for a normal 18 month surveillance interval considering three failures is 8.01E-07 failures per hour which is slightly higher than the generic value of 6.30E-07 failures per hour. For a ten year interval, TVA calculated an upper limit value of 2.39E-06 EFCV failures per hour approximately four (4) times the 1 Operating time is calculated as follows: 67 valves
- 11 cycles
- 1.5 years per cycle = 1105.5 years. 1105.5 years
- 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year = 9.68E+06 hours. However, the last 4 cycles were 2 years in duration but 1.5 years used in calculation.
2 Failure rate for the ten year interval is obtained using a chi-square distribution considering 5 times the normal failure rate in the ten year period, 15 failures.
E1-9 generic value. Although the failure rates are greater than the generic value, TVA considers this a low value.
Any future EFCV failure would be evaluated per the BFN corrective action program. Additionally, the BFN 10 CFR 50.65 Maintenance Rule Program, which is contained in Technical Instruction 0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -
10CFR50.65, specifies that no more than two EFCV failures are allowed as monitored over a 24 month rolling interval.
Conclusions The calculated release frequency estimate is 2.47E-04 events per year for ten year surveillance interval. Implementation of this change represents an increase in release frequency of approximately 1.98E-04 events per year from the current release frequency estimate of 4.93E-05 events per year for two year surveillance interval. This increase is not significant, especially since any postulated coolant leakage is within the capability of the reactor coolant makeup systems, and the consequences of such an accident are not expected to lead to a core damage event.
The radiological consequences from an instrument line break were found to be a very small portion of the 10 CFR 50.67 limits. In the unlikely event that an instrument line breaks and the EFCV fails to close, core damage would not be expected to occur and the doses would be less than regulatory limits.
Therefore, it has been concluded, considering the low consequences of a release, the extension of the test interval does not significantly affect the risk to the public associated with the failure of an instrument line and the failure of an EFCV to perform its intended function.
Periodic testing of a representative sample of EFCVs selected on a performance basis will continue to ensure the reliability of these valves. This, along with the plant design, assures that the assumed release rate in the plants safety analysis remains conservative.
E1-10 5.0 REGULATORY SAFETY ANALYSIS The Tennessee Valley Authority (TVA) is submitting an amendment request to license DPR-33 for the Browns Ferry Nuclear Plant (BFN) Unit 1. The current Technical Specifications (TS) requires each reactor instrumentation line excess flow check valve (EFCV) be actuated to the isolation position every 24 months. The proposed amendment will revise the frequency of Surveillance Requirement (SR) 3.6.1.3.8 by testing a representative sample (approximately 20 percent) of EFCV every 24 months, so each EFCV is tested once every 120 months.
5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The current EFCV frequency requires that each reactor instrument line EFCV be tested every 24 months. The EFCVs are designed to automatically close upon excessive differential pressure including failure of the down stream piping or instrument and will reopen when appropriate. This proposed change will allow a reduction in the number of EFCVs that are verified tested every 24 months, to approximately 20 percent of the valves each cycle. BFN and industry operating experience demonstrates high reliability of these valves.
Neither the EFCVs nor their failure is capable of initiating a previously evaluated accident. Therefore, there is no increase in the probability of occurrence of an accident previously evaluated.
The instrument lines going to the Reactor Coolant Pressure boundary with EFCVs installed have flow restricting devices upstream of the EFCV. The consequences of an unisolable failure of an instrument line have been previously evaluated and meet the intent of NRC Safety Guide 11. The offsite exposure has been calculated to be substantially below the limits of 10 CFR 50.67. The total control room Total Effective Dose Equivalent (TEDE) doses are less than the 5 REM limit and the offsite TEDE doses are less than 10% of the 25 REM limit. Additionally, coolant lost from such a break is inconsequential compared to the makeup capabilities of
E1-11 normal and emergency makeup systems. Although not expected to occur as a result of this change, the affects of a postulated failure of an EFCV to isolate and instrument line break as a result of reduced testing are bounded by TVA analysis.
Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed reduction in EFCV test frequency is bounded by previous evaluation of a line rupture. The proposed change does not introduce new equipment, which could create a new or different kind of accident. No new external threats, release pathways, or equipment failure modes are created. Therefore, the implementation of the proposed change will not create a possibility for an accident of a new or different type than those previously evaluated.
3.0 Does the proposed change involve a significant reduction in a margin of safety?
Response: No The consequences of an unisolable rupture of an instrument line have been previously evaluated and meet the intent NRC Safety Guide 11. The proposed change does not involve a significant reduction in a margin of safety. Therefore, the proposed revised surveillance frequency does not adversely affect the public health and safety, and does not involve any significant safety hazards.
Based on the above, TVA concludes that the proposed amendments present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.
E1-12 5.2 Applicable Regulatory Requirements/Criteria As discussed in Regulatory Guide 1.11, "Instrument Lines Penetrating Primary Reactor Containment," the use of EFCVs satisfies the requirements of General Design Criteria 55 and 56 for automatic isolation capability of lines penetrating containment, while maintaining a highly reliable capability to monitor important parameters inside containment.
The BFN, Units 1, 2 and 3, 10 CFR Part 50.65 Maintenance Rule Program will monitor EFCV reliability. No more than two EFCV failures are allowed as monitored over a 24 month rolling interval.
The control room doses due to an instrument line break will be less than the 10 CFR 50 Appendix A, GDC 19 limits. The off-site doses will be less than 10 percent of the 10 CFR 100 limits. The total control room Total Effective Dose Equivalent (TEDE) doses are less than the 5 rem limit and the offsite TEDE doses are less than 10% of the 25 rem limit (10CFR50.67).
In conclusion, based on the considerations discussed above, (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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.
6.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
E1-13
7.0 REFERENCES
- 1.
NRC memorandum, W.D. Beckner to A.R. Pietrangel, regarding Technical Specification Task Force Travelers, October 31, 2000.
- 2.
NRC letter, S. A. Richards to W. G. Warren (BWROG), Safety Evaluation of General Electric Nuclear Energy Topical Report B21-00658-01, Excess Flow Check Valve Testing Relaxation (TAC Nos. MA7884 and M84809), March 14, 2000.
- 3.
TVA letter, T.E. Abney to NRC, Browns Ferry Nuclear Plant (BFN) - Units 2 and 3 - Technical Specifications Change (TS)
No. 400 - Request for License Amendment - Relaxation of Excess Flow Check Valve (EFCV) Surveillance Testing Frequency, August 11, 2000.
- 4.
NRC letter, W.O. Long to J.A. Scalice, Browns Ferry Units 2 and 3, Proposed Technical Specifications Change Regarding Excess Flow Check Valves, Request for Additional Information (TAC Nos. MA6407 and MA6409), October 5, 2000.
- 5.
TVA letter, T.E. Abney to NRC, Browns Ferry Nuclear Plant (BFN) - Response to Request for Additional Information (RAI)
Regarding Units 2 and 3 Technical Specification (TS) Change TS No. 400 - Request for License Amendment Relaxation of Excess Flow Check Valve (EFCV) Surveillance Testing Frequency (TAC Nos. MA6407 and MA6409), October 20, 2000.
- 6.
NRC letter, W.O. Long to J.A. Scalice, Browns Ferry Nuclear Plant, Units 2 and 3 - Issuance of Amendments Regarding Excess Flow Check Valve Surveillance Intervals (TAC Nos.
MA6407 and MA6409), January 29, 2001.
Browns Ferry Nuclear Plant (BFN)
Technical Specification Change 438 Revision to Excess Flow Check Valve (EFCV)
Surveillance Testing Frequency Proposed Technical Specification Changes (mark-up)
PCIVs 3.6.1.3 BFN-UNIT 1 3.6-16 Amendment No. 234 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.5 Verify the isolation time of each power operated, automatic PCIV, except for MSIVs, is within limits.
In accordance with the Inservice Testing Program SR 3.6.1.3.6 Verify the isolation time of each MSIV is 3 seconds and 5 seconds.
In accordance with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal.
18 months SR 3.6.1.3.8 Verify each a representative sample of reactor instrumentation line EFCVs actuates to the isolation position on a simulated instrument line break signal.
24 months SR 3.6.1.3.9 Remove and test the explosive squib from each shear isolation valve of the TIP System.
18 months on a STAGGERED TEST BASIS SR 3.6.1.3.10 Verify leakage rate through each MSIV is 11.5 scfh when tested at 25 psig.
In accordance with the Primary Containment Leakage Rate Testing Program SR 3.6.1.3.11 Verify combined leakage through water tested lines that penetrate primary containment are within the limits specified in the Primary Containment Leakage Rate Testing Program.
In accordance with the Primary Containment Leakage Rate Testing Program
Browns Ferry Nuclear Plant (BFN)
Technical Specification Change 438 Revision to Excess Flow Check Valve (EFCV)
Surveillance Testing Frequency Proposed Technical Specification Changes (updated pages)
PCIVs 3.6.1.3 BFN-UNIT 1 3.6-16 Amendment No. 234 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.5 Verify the isolation time of each power operated, automatic PCIV, except for MSIVs, is within limits.
In accordance with the Inservice Testing Program SR 3.6.1.3.6 Verify the isolation time of each MSIV is 3 seconds and 5 seconds.
In accordance with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal.
18 months SR 3.6.1.3.8 Verify a representative sample of reactor instrumentation line EFCVs actuates to the isolation position on a simulated instrument line break signal.
24 months SR 3.6.1.3.9 Remove and test the explosive squib from each shear isolation valve of the TIP System.
18 months on a STAGGERED TEST BASIS SR 3.6.1.3.10 Verify leakage rate through each MSIV is 11.5 scfh when tested at 25 psig.
In accordance with the Primary Containment Leakage Rate Testing Program SR 3.6.1.3.11 Verify combined leakage through water tested lines that penetrate primary containment are within the limits specified in the Primary Containment Leakage Rate Testing Program.
In accordance with the Primary Containment Leakage Rate Testing Program