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| | number = ML061650267 | | | number = ML061650267 |
| | issue date = 06/12/2006 | | | issue date = 06/12/2006 |
| | title = Watts Bar Nuclear Plant (WBN) Unit 1 - Technical Specifications (TS) Change WBN-TS-05-10 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity - Request for Additional Information (TAC MC9271) | | | title = Technical Specifications (TS) Change WBN-TS-05-10 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity - Request for Additional Information |
| | author name = Pace P L | | | author name = Pace P |
| | author affiliation = Tennessee Valley Authority | | | author affiliation = Tennessee Valley Authority |
| | addressee name = | | | addressee name = |
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| | page count = 10 | | | page count = 10 |
| | project = TAC:MC9271 | | | project = TAC:MC9271 |
| | stage = RAI | | | stage = Request |
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| {{#Wiki_filter:Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 JUN 12 2006 WBN-TS-05-10 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen: | | {{#Wiki_filter:Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 JUN 12 2006 WBN-TS-05-10 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen: |
| In the Matter of ) Docket No. 50-390 Tennessee Valley Authority WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 -TECHNICAL SPECIFICATIONS (TS) CHANGE WBN-TS-05-10 | | In the Matter of ) Docket No. 50-390 Tennessee Valley Authority WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - TECHNICAL SPECIFICATIONS (TS) CHANGE WBN-TS-05 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY - REQUEST FOR ADDITIONAL INFORMATION (TAC NO. MC 9271) |
| -APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY
| | The purpose of this letter is to provide TVA's response to the request for additional information dated May 9, 2006, concerning the subject amendment request that was submitted to NRC on December 15, 2005. provides TVA's response to NRC's questions. There are no regulatory commitments associated with this submittal. |
| -REQUEST FOR ADDITIONAL INFORMATION (TAC NO. MC 9271)The purpose of this letter is to provide TVA's response to the request for additional information dated May 9, 2006, concerning the subject amendment request that was submitted to NRC on December 15, 2005.Enclosure 1 provides TVA's response to NRC's questions. | | If you have any questions concerning this matter, please call me at (423) 365-1824. |
| There are no regulatory commitments associated with this submittal. | | Printed on recycled paper |
| If you have any questions concerning this matter, please call me at (423) 365-1824.Printed on recycled paper U.S. Nuclear Regulatory Commission Page 2 JUN 12 2006 I declare under penalty of perjury that the foregoing is true and correct. Executed on this 1 2 th day of June 2006.Sincerely, P. L. Pace Manager, Site Licensing and Industry Affairs Enclosures | | |
| : 1. Response to RAI Questions 2. Revised Technical Specification Page 3. Revised Technical Specification Bases Pages cc (Enclosures): | | U.S. Nuclear Regulatory Commission Page 2 JUN 12 2006 I declare under penalty of perjury that the foregoing is true and correct. Executed on this 1 2 th day of June 2006. |
| NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY | | Sincerely, P. L. Pace Manager, Site Licensing and Industry Affairs Enclosures |
| -TSTF-449, REVISION 4 TVA submitted an application for an amendment to revise the WBN Unit 1 technical specification (TS) requirements to be consistent with Technical Specification Task Force (TSTF) Traveler, TSTF-449, Revision 4, "Steam Generator Tube Integrity," by letter dated December 15, 2005. As stated on page E1-3 of Enclosure 1 to the application, the current steam generators will be replaced in the Fall 2006. In addition, the approved alternate repair criteria (ARC) (i.e., voltage based ARC for outside diameter stress corrosion cracking and the use of the F-star), and the sleeving repair method will be deleted as part of this TS change.NRC issued a request for additional information concerning the subject TS change dated May 9, 2006. TVA's response is provided below: NRC QUESTION 1 Insert A of Enclosure 2 to the application contains TS 5.7.2.12,"Steam Generator (SG) Program," and corresponds to Insert 5.5.9 of TSTF-449, Revision 4.The last sentence in TS 5.7.2.12.a states, "Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected and/or plugged, to confirm that the performance criteria are being met." The intent of this paragraph is to ensure that condition monitoring assessments are conducted when the SG tubes are inspected or plugged as stated in paragraph a of Insert 5.5.9 of TSTF-449, Revision 4. The staff requests the licensee to either justify the use of "and/or" in the last sentence of TS 5.7.2.12.a or to replace "and/or" with"or.RESPONSE See attached change to TS 5.7.2.12.a. | | : 1. Response to RAI Questions |
| -"and/or" has been replaced with "or." NRC QUESTION 2 Insert A of Enclosure 2 to the application contains TS 5.7.2.12,"Steam Generator (SG) Program," and corresponds to Insert 5.5.9 of TSTF-449, Revision 4.The last sentence in TS 5.7.2.12.b.2 states, "The accident induced leakage is not to exceed 1.0 gpm for the faulted SG. " The corresponding sentence in Insert 5.5.9 of TSTF-449, Revision EI-l ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY | | : 2. Revised Technical Specification Page |
| -TSTF-449, REVISION 4 4, states, "Leakage is not to exceed [1 gpm] per SG [, except for specific types of degradation at specific locations described in paragraph c of the Steam Generator Program]." The intent of this sentence is to ensure that leakage does not exceed 1 gpm in any SG except for those instances defined in paragraph c of the SG program. The staff requests the licensee to discuss its plans to modify this sentence consistent with TSTF-449, Revision 4.RESPONSE TVA acknowledges that the TSTF, Insert 5.5.9 states "per SG." However, the use of the words "per SG" indicates WBN could have a total from all four steam generators of four gallons per minute (gpm) primary-to-secondary leakage during an Main Steam Line Break (MSLB) accident. | | : 3. Revised Technical Specification Bases Pages cc (Enclosures): |
| The MSLB dose calculation for the replacement steam generators uses one gpm in the faulted steam generator and 150 gallons per day (gpd) which is 0.10 gpm in the non-faulted steam generators (See Technical Specification Bases 3.4.13, RCS Operational Leakage, under Applicable Safety Analysis, page B 3.4-75, contained in TVA's request, WBN-TS-05-10, dated December 15, 2005). Therefore, TVA considers the wording change justified to prevent a misinterpretation which would be outside the WBN design basis.NRC QUESTION 3 Enclosure 3 to the application includes the new bases for TS 3.4.17, "Steam Generator (SG) Tube Integrity," and corresponds to section B 3.4.20, "Steam Generator (SG) Tube Integrity," found on pages B 3.4.20-1 through B 3.4.20-7 of TSTF-449, Revision 4.On page B 3.4-100 of the application, second paragraph under"Applicable Safety Analysis," the last sentence reads, "The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3)." The corresponding sentence on page B 3.4.20-2 of TSTF-449, Revision 4, includes, "or the NRC approved licensing basis (e.g., a small fraction of these limits)." The staff requests the licensee to justify the exclusion of this phrase from its application. | | NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 |
| RESPONSE The calculated Main Control Room thyroid dose for the replacement steam generators in the MSLB radiological dose calculation is 12.5 rem versus the limit of 30 rem. Therefore, TVA did not include the words in parenthesis that state "(e.g., a small EI-2 ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY | | |
| -TSTF-449, REVISION 4 fraction of these limits)," as 12.5 rem can not be considered a small fraction of 30 rem. However, 12.5 rem is well within the NRC's regulatory limits. TVA has added the words to the end of Bases second paragraph under Applicable Safety Analyses, page B 3.4-100, "or the NRC approved licensing basis." See attached mark-up of page B 3.4-100.NRC QUESTION 4 Enclosure 3 to the application includes the new bases for TS 3.4.17, "Steam Generator (SG) Tube Integrity," and corresponds to section B 3.4.20, "Steam Generator (SG) Tube Integrity," found on pages B 3.4.20-1 through B 3.4.20-7 of TSTF-449, Revision 4.On page B 3.4-102 of the application, second paragraph under"Actions," the last sentence reads, "If it is determined that tube integrity is not being maintained until the next SG inspection, Condition B applies." The corresponding sentence on page B.3.4.20-4 of TSTF-449, Revision 4, does not include the phrase "until the next SG inspection," however, the required action found in Enclosure 2 of the application, page 3.4-43, includes the phrase, "until the next refueling outage or SG tube inspection." The staff requests the licensee to justify the deviation from the action described in TS 3.4.17 and the wording in TSTF-449, Revision 4.RESPONSE See attached change to B 3.4-102. "..until the next SG inspection," has been deleted to be consistent with the TSTF.EI-3 ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY | | ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 TVA submitted an application for an amendment to revise the WBN Unit 1 technical specification (TS) requirements to be consistent with Technical Specification Task Force (TSTF) Traveler, TSTF-449, Revision 4, "Steam Generator Tube Integrity," by letter dated December 15, 2005. As stated on page E1-3 of Enclosure 1 to the application, the current steam generators will be replaced in the Fall 2006. In addition, the approved alternate repair criteria (ARC) (i.e., voltage based ARC for outside diameter stress corrosion cracking and the use of the F-star), and the sleeving repair method will be deleted as part of this TS change. |
| -TSTF-449, REVISION 4 REVISED PROPOSED TECHNICAL SPECIFICATION PAGES (MARK UP)I. Affected Page List 5.0-15 Insert A II. Marked Pages See Attached INSERT A 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Proaram A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. | | NRC issued a request for additional information concerning the subject TS change dated May 9, 2006. TVA's response is provided below: |
| In addition, the Steam Generator Program shall include the following provisions: | | NRC QUESTION 1 Insert A of Enclosure 2 to the application contains TS 5.7.2.12, "Steam Generator (SG) Program," and corresponds to Insert 5.5.9 of TSTF-449, Revision 4. |
| : a. Provisions for condition monitoring assessments. | | The last sentence in TS 5.7.2.12.a states, "Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected and/or plugged, to confirm that the performance criteria are being met." The intent of this paragraph is to ensure that condition monitoring assessments are conducted when the SG tubes are inspected or plugged as stated in paragraph a of Insert 5.5.9 of TSTF-449, Revision 4. The staff requests the licensee to either justify the use of "and/or" in the last sentence of TS 5.7.2.12.a or to replace "and/or" with "or. |
| Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.b. Performance criteria for SG tube integrity. | | |
| SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion: | | ===RESPONSE=== |
| All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents. | | See attached change to TS 5.7.2.12.a. - "and/or" has been replaced with "or." |
| This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. | | NRC QUESTION 2 Insert A of Enclosure 2 to the application contains TS 5.7.2.12, "Steam Generator (SG) Program," and corresponds to Insert 5.5.9 of TSTF-449, Revision 4. |
| Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. | | The last sentence in TS 5.7.2.12.b.2 states, "The accident induced leakage is not to exceed 1.0 gpm for the faulted SG. " |
| In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.2. Accident induced leakage performance criterion: | | The corresponding sentence in Insert 5.5.9 of TSTF-449, Revision EI-l |
| The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. The accident induced leakage is not to exceed 1.0 gpm for the faulted SG.3. The operational leakage performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE." c. Provisions for SG tube repair criteria. | | |
| Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.d. Provisions for SG tube inspections. | | ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 4, states, "Leakage is not to exceed [1 gpm] per SG [, except for specific types of degradation at specific locations described in paragraph c of the Steam Generator Program]." The intent of this sentence is to ensure that leakage does not exceed 1 gpm in any SG except for those instances defined in paragraph c of the SG program. The staff requests the licensee to discuss its plans to modify this sentence consistent with TSTF-449, Revision 4. |
| Periodic SG tube inspections shall be performed. | | |
| The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. | | ===RESPONSE=== |
| The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. | | TVA acknowledges that the TSTF, Insert 5.5.9 states "per SG." |
| An assessment of degradation shall be performed to determine the type and ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY | | However, the use of the words "per SG" indicates WBN could have a total from all four steam generators of four gallons per minute (gpm) primary-to-secondary leakage during an Main Steam Line Break (MSLB) accident. The MSLB dose calculation for the replacement steam generators uses one gpm in the faulted steam generator and 150 gallons per day (gpd) which is 0.10 gpm in the non-faulted steam generators (See Technical Specification Bases 3.4.13, RCS Operational Leakage, under Applicable Safety Analysis, page B 3.4-75, contained in TVA's request, WBN-TS-05-10, dated December 15, 2005). Therefore, TVA considers the wording change justified to prevent a misinterpretation which would be outside the WBN design basis. |
| -TSTF-449, REVISION 4 REVISED PROPOSED TECHNICAL SPECIFICATION PAGES I. Affected Page List B 3.4-100 B 3.4-102 II. Marked Pages See Attached SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY ANALYSES basis event for SG tubes and avoiding an SGTR is the basis for this Specification. | | NRC QUESTION 3 to the application includes the new bases for TS 3.4.17, "Steam Generator (SG) Tube Integrity," and corresponds to section B 3.4.20, "Steam Generator (SG) Tube Integrity," found on pages B 3.4.20-1 through B 3.4.20-7 of TSTF-449, Revision 4. |
| The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser. | | On page B 3.4-100 of the application, second paragraph under "Applicable Safety Analysis," the last sentence reads, "The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3)." The corresponding sentence on page B 3.4.20-2 of TSTF-449, Revision 4, includes, "or the NRC approved licensing basis (e.g., a small fraction of these limits)." The staff requests the licensee to justify the exclusion of this phrase from its application. |
| The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). | | |
| In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from 150 gallons per day (gpd) per steam generator and 1 gallon per minute (gpm) in the faulted steam generator. | | ===RESPONSE=== |
| For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16 "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3) or the NRC approved licensing basis.Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). | | The calculated Main Control Room thyroid dose for the replacement steam generators in the MSLB radiological dose calculation is 12.5 rem versus the limit of 30 rem. Therefore, TVA did not include the words in parenthesis that state "(e.g., a small EI-2 |
| LCO The LCO requires that SG tube integrity be maintained. | | |
| The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. | | ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 fraction of these limits)," as 12.5 rem can not be considered a small fraction of 30 rem. However, 12.5 rem is well within the NRC's regulatory limits. TVA has added the words to the end of Bases second paragraph under Applicable Safety Analyses, page B 3.4-100, "or the NRC approved licensing basis." See attached mark-up of page B 3.4-100. |
| If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity. | | NRC QUESTION 4 to the application includes the new bases for TS 3.4.17, "Steam Generator (SG) Tube Integrity," and corresponds to section B 3.4.20, "Steam Generator (SG) Tube Integrity," found on pages B 3.4.20-1 through B 3.4.20-7 of TSTF-449, Revision 4. |
| In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.A SG tube has tube integrity when it satisfies the SG performance criteria.The SG performance criteria are defined in Specification 5.7.2.12, "Steam Generator Program," and describe acceptable SG tube performance. | | On page B 3.4-102 of the application, second paragraph under "Actions," the last sentence reads, "If it is determined that tube integrity is not being maintained until the next SG inspection, Condition B applies." The corresponding sentence on page B.3.4.20-4 of TSTF-449, Revision 4, does not include the phrase "until the next SG inspection," however, the required action found in Enclosure 2 of the application, page 3.4-43, includes the phrase, "until the next refueling outage or SG tube inspection." The staff requests the licensee to justify the deviation from the action described in TS 3.4.17 and the wording in TSTF-449, Revision 4. |
| The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.(continued) | | |
| Watts Bar-Unit 1 B 3.4-100 Revision Amendment SG Tube Integrity B 3.4.17 BASES LCO (continued) | | ===RESPONSE=== |
| The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. | | See attached change to B 3.4-102. "..until the next SG inspection," has been deleted to be consistent with the TSTF. |
| The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day.This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative. | | EI-3 |
| APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.ACTIONS The ACTIONS are modified by a Note that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube.Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry, and application of associated Required Actions.A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. | | |
| An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged, has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. | | ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 REVISED PROPOSED TECHNICAL SPECIFICATION PAGES (MARK UP) |
| The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. | | I. Affected Page List 5.0-15 Insert A II. Marked Pages See Attached |
| If it is determined that tube integrity is not being maintained until the next SG.....ie~I, -Condition B applies.(continued) | | |
| I Watts Bar-Unit I B 3.4-102 Revision Amendment}}
| | INSERT A 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Proaram A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions: |
| | : a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met. |
| | : b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE. |
| | : 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. |
| | Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. |
| | : 2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. The accident induced leakage is not to exceed 1.0 gpm for the faulted SG. |
| | : 3. The operational leakage performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE." |
| | : c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged. |
| | : d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and |
| | |
| | ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 REVISED PROPOSED TECHNICAL SPECIFICATION PAGES I. Affected Page List B 3.4-100 B 3.4-102 II. Marked Pages See Attached |
| | |
| | SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY ANALYSES basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser. |
| | The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from 150 gallons per day (gpd) per steam generator and 1 gallon per minute (gpm) in the faulted steam generator. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16 "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3) or the NRC approved licensing basis. |
| | Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). |
| | LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program. |
| | During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity. |
| | In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube. |
| | A SG tube has tube integrity when it satisfies the SG performance criteria. |
| | The SG performance criteria are defined in Specification 5.7.2.12, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria. |
| | (continued) |
| | Watts Bar-Unit 1 B 3.4-100 Revision Amendment |
| | |
| | SG Tube Integrity B 3.4.17 BASES LCO The operational LEAKAGE performance criterion provides an observable (continued) indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. |
| | This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative. |
| | APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4. |
| | RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE. |
| | ACTIONS The ACTIONS are modified by a Note that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. |
| | Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry, and application of associated Required Actions. |
| | A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged, has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained until the next SG |
| | ..... ie~I, |
| | -Condition B applies. I (continued) |
| | Watts Bar-Unit I B 3.4-102 Revision Amendment}} |
Letter Sequence Request |
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TAC:MC9271, Steam Generator Tube Integrity (Approved, Closed) |
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MONTHYEARML0535400672005-12-15015 December 2005 Technical Specifications (TS) Change WBN-TS-05-10 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity Project stage: Request ML0608904642006-05-0909 May 2006 Request for Additional Information Regarding Steam Generator Tube Integrity (TSTF-449) Project stage: RAI ML0616502672006-06-12012 June 2006 Technical Specifications (TS) Change WBN-TS-05-10 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity - Request for Additional Information Project stage: Request ML0621800752006-08-10010 August 2006 RAI Re Steam Generator Tube Integrity (TSTF-449) Project stage: RAI ML0625503712006-09-0808 September 2006 Technical Specifications (TS) Change WBN-TS-05-10 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity - Request for Additional Information Project stage: Request ML0627202242006-10-10010 October 2006 Staff'S Review of the Steam Generator Tube Integrity Technical Specification Amendment Project stage: Approval ML0631103212006-11-0303 November 2006 Tech Spec Pages for Amendment 65 Regarding Steam Generator Tube Integrity (TS-05-10) Project stage: Acceptance Review ML0629100932006-11-0303 November 2006 License Amendment 65 Regarding Steam Generator Tube Integrity (TS-05-10) Project stage: Approval 2006-05-09
[Table View] |
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Category:Letter
MONTHYEARML23319A2452024-01-29029 January 2024 Issuance of Amendment Nos. 366 and 360; 164 and 71 Regarding the Adoption of TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues ML24008A2462024-01-18018 January 2024 Revision to the Reactor Vessel Material Surveillance Capsule Withdrawal Schedule CNL-24-017, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions2024-01-17017 January 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure and Emergency Preparedness Department Procedure Revisions CNL-24-016, Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14)2024-01-10010 January 2024 Supplement to Application to Modify the Technical Specification Surveillance Requirement 3.9.5.1 (WBN-TS-21-14) CNL-23-052, Application to Adopt TSTF-427-A, Revision 2, Allowance for Non-Technical Specification Barrier Degradation on Supported System Operability2024-01-0909 January 2024 Application to Adopt TSTF-427-A, Revision 2, Allowance for Non-Technical Specification Barrier Degradation on Supported System Operability CNL-23-062, Application to Revise the Technical Specifications Section 3.8.2, AC Sources-Shutdown, to Remove Reference to the C-S Diesel Generator (WBN-TS-23-018)2024-01-0808 January 2024 Application to Revise the Technical Specifications Section 3.8.2, AC Sources-Shutdown, to Remove Reference to the C-S Diesel Generator (WBN-TS-23-018) ML23346A1382024-01-0303 January 2024 Regulatory Audit Summary Related to Request to Increase the Number of Tritium Producing Burnable Absorber Rods CNL-23-069, Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation2023-12-21021 December 2023 Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation IR 05000390/20234412023-12-21021 December 2023 Plantfinal Significance Determination for a Security-Related Greater than Green Finding, Nov, and Assessment Follow-up, 05000390-2023441 and 05000391-2023441-Public CNL-23-036, Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08)2023-12-18018 December 2023 Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08) IR 05000390/20234042023-12-14014 December 2023 Security Baseline Inspection Report 05000390/2023404 and 05000391/2023404 CNL-23-001, Rebaseline of Sections 3.1 and 3.2 of the Technical Specifications (WBN-TS-23-01)2023-12-13013 December 2023 Rebaseline of Sections 3.1 and 3.2 of the Technical Specifications (WBN-TS-23-01) ML23293A0572023-12-0606 December 2023 Issuance of Amendment Nos. 163 and 70 Regarding Adoption of TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control IR 05000390/20230102023-11-30030 November 2023 RE-Issue Watts Bar Nuclear Plant - Biennial Problem Identification and Resolution Inspection Report 050000390/2023010 and 05000391/2023010 and Apparent Violation CNL-23-067, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-11-27027 November 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000390/20230032023-11-13013 November 2023 Integrated Inspection Report 05000390/2023003 and 05000391/2023003 and Apparent Violation ML23312A1432023-11-0808 November 2023 Submittal of Dual Unit Updated Final Safety Analysis Report (UFSAR) Amendment 5 CNL-23-059, Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-2023-09-20020 September 2023 Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 CNL-23-061, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-09-20020 September 2023 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revision ML23251A2002023-09-11011 September 2023 Request for Withholding Information from Public Disclosure for Watts Bar Nuclear Plant, Units 1 and 2 CNL-23-057, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 642023-09-0505 September 2023 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions. Includes CECC-EPIP-1, Revision 76 and CECC-EPIP-9, Revision 64 IR 05000390/20230052023-08-30030 August 2023 Updated Inspection Plan for Watts Bar Nuclear Plant, Units 1 and 2 - Report 05000390/2023005 and 05000391/2023005 ML23233A0042023-08-28028 August 2023 Proposed Alternative to the Requirements of the ASME Boiler and Pressure Vessel Code for Upper Head Injection Dissimilar Metal Butt Welds IR 05000390/20230022023-08-16016 August 2023 Reissue - Watts Bar Nuclear Plant - Integrated Inspection Report 05000390/2023002 and 05000391/2023002 ML23220A1582023-08-0909 August 2023 Integrated Inspection Report 05000390/2023002 and 05000391/2023002 CNL-23-045, License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts (WBN-TS-23-010)2023-08-0707 August 2023 License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts (WBN-TS-23-010) CNL-23-028, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06)2023-08-0202 August 2023 Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06) ML23192A4472023-07-31031 July 2023 Staff Assessment of Updated Seismic Hazards at TVA Sites Following the NRC Process for the Ongoing Assessment of Natural Hazards Information CNL-23-055, Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills2023-07-25025 July 2023 Tennessee Valley Authority - Emergency Preparedness Department Procedure Revision. Includes EPDP-3, Revision 3, Emergency Plan Exercises and Preparedness Drills IR 05000390/20230112023-07-24024 July 2023 Quadrennial Focused Engineering Inspection (FEI) Commercial Grade Dedication Report 05000390 2023011 and 05000391 2023011 CNL-23-053, Tennessee Valley Authority - Radiological Emergency Plan Revisions2023-07-18018 July 2023 Tennessee Valley Authority - Radiological Emergency Plan Revisions CNL-23-020, Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06)2023-06-28028 June 2023 Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06) CNL-23-049, Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan .2023-06-26026 June 2023 Tennessee Valley Authority - Radiological Emergency Plan and Central Emergency Control Center Emergency Plan . ML23122A2322023-06-0707 June 2023 Issuance of Amendment Nos. 162 and 69 Regarding Change to Date in Footnotes for Technical Specification 3.7.11, Control Room Emergency Air Temperature Control System (Creatcs) CNL-23-044, Transmittal of Revision 3 to WCAP-18774-P and WCAP-18774-NP, Addendum to the Rotterdam Dockyard Company Final Stress Report for 173 P.W.R. Vessels TVA III & IV (Report No. 30749-B-030, Rev. 3) - Evaluation of One Closure Stud Out2023-06-0101 June 2023 Transmittal of Revision 3 to WCAP-18774-P and WCAP-18774-NP, Addendum to the Rotterdam Dockyard Company Final Stress Report for 173 P.W.R. Vessels TVA III & IV (Report No. 30749-B-030, Rev. 3) - Evaluation of One Closure Stud Out IR 05000390/20234032023-05-30030 May 2023 Cyber Security Inspection Report 05000390/2023403 and 05000391/2023403 ML23131A1812023-05-23023 May 2023 Correction to Amendment No. 161 to Facility Operating License No. NPF-90 CNL-23-042, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-05-16016 May 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000390/20220032023-05-0909 May 2023 Reissue Watts Bar Nuclear Plant - Integrated Inspection Report 05000390/2022003 and 05000391/2022003 ML23125A2202023-05-0505 May 2023 Issuance of Amendment No. 161 Regarding a Change to Footnotes for Technical Specification Table 1.1-1 Modes (Emergency Circumstances) IR 05000390/20230012023-05-0404 May 2023 Integrated Inspection Report 05000390/2023001 and 05000391/2023001 CNL-23-043, Emergency License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes (WBN-TS-23-09)2023-05-0404 May 2023 Emergency License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes (WBN-TS-23-09) CNL-23-032, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 412023-04-27027 April 2023 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 41 CNL-23-030, Tennessee Valley Authority - Hydrologic Engineering Center River Analysis System Project Milestone Status Update2023-04-27027 April 2023 Tennessee Valley Authority - Hydrologic Engineering Center River Analysis System Project Milestone Status Update CNL-23-033, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revision2023-04-24024 April 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revision CNL-23-029, Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2023-04-11011 April 2023 Radiological Emergency Plan and Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML23072A0652023-04-0505 April 2023 Units 1 and 2 Issuance of Amendment Nos. 364 and 358; 160 and 68 Regarding a Revision to Technical Specification 3.4.12 ML23073A2762023-04-0303 April 2023 Individual Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing (EPID L-2023-LLA-0029) (Letter) CNL-23-023, Annual Insurance Status Report2023-03-30030 March 2023 Annual Insurance Status Report CNL-23-024, TVA Guarantee of Payment of Deferred Premiums - 2022 Annual Report2023-03-29029 March 2023 TVA Guarantee of Payment of Deferred Premiums - 2022 Annual Report 2024-01-09
[Table view] Category:License-Application for Facility Operating License (Amend/Renewal) DKT 50
MONTHYEARCNL-23-036, Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08)2023-12-18018 December 2023 Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08) CNL-23-059, Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-2023-09-20020 September 2023 Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 CNL-23-045, License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts (WBN-TS-23-010)2023-08-0707 August 2023 License Amendment Request to Revise Technical Specification Table 1.1-1 Regarding the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Bolts (WBN-TS-23-010) CNL-23-028, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06)2023-08-0202 August 2023 Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06) CNL-23-043, Emergency License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes (WBN-TS-23-09)2023-05-0404 May 2023 Emergency License Amendment Request to Relax the Required Number of Fully Tensioned Reactor Pressure Vessel Head Closure Studs in Technical Specification Table 1.1-1, Modes (WBN-TS-23-09) CNL-23-002, Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Sche2023-03-20020 March 2023 Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Schedu CNL-23-015, Expedited Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-22-08)2023-02-27027 February 2023 Expedited Application to Modify the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-22-08) CNL-22-030, Application to Revise Technical Specification 3.4.12, Low Temperature Overpressure Protection System for Sequoyah Nuclear Plant (SQN-TSC-22-01) and TS 3.4.12 Cold Overpressure Mitigation System for Watts Bar Nuclear Plant (WBN-TS-22-03)2022-07-27027 July 2022 Application to Revise Technical Specification 3.4.12, Low Temperature Overpressure Protection System for Sequoyah Nuclear Plant (SQN-TSC-22-01) and TS 3.4.12 Cold Overpressure Mitigation System for Watts Bar Nuclear Plant (WBN-TS-22-03) CNL-22-071, Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections (SQN-TS-21-03 and WBN-TS-21-08)2022-07-13013 July 2022 Supplement to Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections (SQN-TS-21-03 and WBN-TS-21-08) CNL-22-039, Application to Revise Technical Specifications to Adopt TSTF-554-A, Revision 1, Revise Reactor Coolant Leakage Requirements (BFN TS-537) (SQN-21-05) (WBN-TS-21-04)2022-07-13013 July 2022 Application to Revise Technical Specifications to Adopt TSTF-554-A, Revision 1, Revise Reactor Coolant Leakage Requirements (BFN TS-537) (SQN-21-05) (WBN-TS-21-04) CNL-22-008, and Watts Bar Nuclear Plant - Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules (BFN-TS-541, SQN-TS-21-09, and WBN-TS-22-002)2022-06-13013 June 2022 and Watts Bar Nuclear Plant - Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules (BFN-TS-541, SQN-TS-21-09, and WBN-TS-22-002) CNL-22-001, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections (SQN-TS-21-03 and WBN-TS-21-08)2022-04-0404 April 2022 Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections (SQN-TS-21-03 and WBN-TS-21-08) CNL-22-003, Application to Modify the Allowable Value for Watts Bar Nuclear Plant, Unit 1 Technical Specification Table 3.3.2-1, Function 6.e(1) (WBN-TS-21-010)2022-02-17017 February 2022 Application to Modify the Allowable Value for Watts Bar Nuclear Plant, Unit 1 Technical Specification Table 3.3.2-1, Function 6.e(1) (WBN-TS-21-010) CNL-21-018, Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance .2021-12-0909 December 2021 Application to Adopt TSTF-205-A, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance . CNL-21-062, Application to Modify Watts Bar Nuclear Plant Units 1 and 2 Technical Specification 3.7.8 to Support Shutdown Board Cleaning (WBN-TS-19-019)2021-09-29029 September 2021 Application to Modify Watts Bar Nuclear Plant Units 1 and 2 Technical Specification 3.7.8 to Support Shutdown Board Cleaning (WBN-TS-19-019) ML21202A2242021-07-0202 July 2021 Tennessee Valley Authority (TVA)- Watts Bar Nuclear Plant (WBN)- NPDES Permit No. TN0020168-Application Request for Permit Renewal CNL-21-038, Application to Revise the Measurement Units for the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for the Containment Vent Isolation Instrumentation and Control Room Emergency Ventilation System Radiation Monitors (W2021-06-0101 June 2021 Application to Revise the Measurement Units for the Watts Bar Nuclear Plant, Unit 1 and Unit 2 Technical Specifications for the Containment Vent Isolation Instrumentation and Control Room Emergency Ventilation System Radiation Monitors (WBN CNL-21-045, Bellefonte Nuclear Plant, Units 1 and 2; Browns Ferry Nuclear Plant, Units 1, 2, and 3; Clinch River Nuclear Site; Sequoyah Nuclear Plant, Units 1 and 2; Watts Bar Nuclear Plant, Unit 1 and 2 - Nuclear Quality Assurance Plan, TVA-NQA-PLN82021-04-29029 April 2021 Bellefonte Nuclear Plant, Units 1 and 2; Browns Ferry Nuclear Plant, Units 1, 2, and 3; Clinch River Nuclear Site; Sequoyah Nuclear Plant, Units 1 and 2; Watts Bar Nuclear Plant, Unit 1 and 2 - Nuclear Quality Assurance Plan, TVA-NQA-PLN89- CNL-21-013, Application to Modify Technical Specifications Steam Generator Inspection/Repair Program Provisions and Unit 2 Facility Operating License Condition 2.C.(4) (WBN-TS-20-06)2021-03-11011 March 2021 Application to Modify Technical Specifications Steam Generator Inspection/Repair Program Provisions and Unit 2 Facility Operating License Condition 2.C.(4) (WBN-TS-20-06) CNL-21-015, Expedited Application to Modify Watts Bar Nuclear Plant, Unit 1 Technical Specification 3.7.12, Auxiliary Building Gas Treatment System, for One-Time Exception to Permit Opening of the Auxiliary Building Secondary Containment Enclosure as2021-03-0303 March 2021 Expedited Application to Modify Watts Bar Nuclear Plant, Unit 1 Technical Specification 3.7.12, Auxiliary Building Gas Treatment System, for One-Time Exception to Permit Opening of the Auxiliary Building Secondary Containment Enclosure as N CNL-21-016, License Amendment Request to Revise Technical Specifications to Change the Steam Generator Secondary Side Water Level to Accommodate the Replacement Steam Generators (WBN-TS-20-05)2021-03-0202 March 2021 License Amendment Request to Revise Technical Specifications to Change the Steam Generator Secondary Side Water Level to Accommodate the Replacement Steam Generators (WBN-TS-20-05) CNL-21-010, Correction of Application to Implement the Full Spectrum LOCA (Fsloca) Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology (WBN-TS-19-04) (EPID2021-01-26026 January 2021 Correction of Application to Implement the Full Spectrum LOCA (Fsloca) Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology (WBN-TS-19-04) (EPID L CNL-20-104, Expedited Application for Approval to Use an Alternate Method of Determining Probability of Detection for (WBN TS-391-20-024)2020-12-23023 December 2020 Expedited Application for Approval to Use an Alternate Method of Determining Probability of Detection for (WBN TS-391-20-024) CNL-20-093, Supplement to Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 'Containment Leakage Rate Testing Program' (WBN-TS-19-01)2020-12-15015 December 2020 Supplement to Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 'Containment Leakage Rate Testing Program' (WBN-TS-19-01) CNL-20-087, Non-Voluntary Application to Modify Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications Surveillance Requirement 3.6.15.4, Shield Building, (WBN-TS-20-017)2020-12-15015 December 2020 Non-Voluntary Application to Modify Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications Surveillance Requirement 3.6.15.4, Shield Building, (WBN-TS-20-017) CNL-20-078, Supplement to Application to Revise Watts Bar Nuclear Plant (Wbn), Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency and to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sa2020-10-13013 October 2020 Supplement to Application to Revise Watts Bar Nuclear Plant (Wbn), Unit 1 Technical Specifications for Steam Generator Tube Inspection Frequency and to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Samp CNL-20-006, Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 Containment Leakage Rate Testing Program (WBN-TS-19-01)2020-10-0202 October 2020 Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications 5.7.2.19 Containment Leakage Rate Testing Program (WBN-TS-19-01) CNL-20-058, Tennessee Valley Authority, Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications for Function 6.e of Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation (WBN-TS-19-27)2020-08-27027 August 2020 Tennessee Valley Authority, Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications for Function 6.e of Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation (WBN-TS-19-27) CNL-20-041, License Amendment Request to Remove Licensee Control (BFN TS-527, SQN-TS-20-08, and WBN-TS-20-016)2020-08-14014 August 2020 License Amendment Request to Remove Licensee Control (BFN TS-527, SQN-TS-20-08, and WBN-TS-20-016) CNL-20-047, Brown Ferry Nuclear Plant, Sequoyah Nuclear Plant & Watts Bar Nuclear Plant - Tennessee Valley Authority License Amendment Request to Revise Radiological Emergency Plan Regarding On-shift Emergency Medical Technician and Onsite Ambulance2020-07-31031 July 2020 Brown Ferry Nuclear Plant, Sequoyah Nuclear Plant & Watts Bar Nuclear Plant - Tennessee Valley Authority License Amendment Request to Revise Radiological Emergency Plan Regarding On-shift Emergency Medical Technician and Onsite Ambulance Re CNL-20-063, Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06)2020-07-27027 July 2020 Supplement to License Amendment Request for Measurement Uncertainty Recapture Power Uprate (WBN-TS-19-06) CNL-20-053, Application to Revise Watts Bar Nuclear Plant, Unit 1, Technical Specifications for Steam Generator Tube Inspection Frequency & to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies & Tube Sample Selection, (WBN-392020-07-17017 July 2020 Application to Revise Watts Bar Nuclear Plant, Unit 1, Technical Specifications for Steam Generator Tube Inspection Frequency & to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies & Tube Sample Selection, (WBN-390- CNL-20-040, License Amendment Request to Adopt TSTF-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec (WBN-TS-20-07)2020-07-17017 July 2020 License Amendment Request to Adopt TSTF-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec (WBN-TS-20-07) CNL-20-016, Application to Modify Watts Bar Nuclear Plant Unit 1 Technical Specifications 3.3.3, Post Accident Monitoring Instrumentation, (WBN-TS-19-23)2020-06-22022 June 2020 Application to Modify Watts Bar Nuclear Plant Unit 1 Technical Specifications 3.3.3, Post Accident Monitoring Instrumentation, (WBN-TS-19-23) CNL-20-012, Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-18-16)2020-05-19019 May 2020 Application to Modify the Watts Bar Nuclear Plant Unit 1 and Unit 2 Technical Specifications for Main Control Room Chiller Completion Time Extension (WBN-TS-18-16) CNL-20-005, License Amendment Request to Adopt TSTF-541, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position, (WBN-TS-20-02)2020-04-0303 April 2020 License Amendment Request to Adopt TSTF-541, Add Exceptions to Surveillance Requirements for Valves and Dampers Locked in the Actuated Position, (WBN-TS-20-02) CNL-19-115, Non-Voluntary License Amendment Request to Modify Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications 3.2.1, Fq(Z), to Implement Methodology from WCAP-17661, Revision 1, Improved RAOC and CAOC Fq Surveillance Technical Specific2020-03-0202 March 2020 Non-Voluntary License Amendment Request to Modify Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications 3.2.1, Fq(Z), to Implement Methodology from WCAP-17661, Revision 1, Improved RAOC and CAOC Fq Surveillance Technical Specificat CNL-19-084, Second Supplement to Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program...2020-01-17017 January 2020 Second Supplement to Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program... CNL-19-051, Application to Implement the Full Spectrum LOCA (Fsloca) Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology2020-01-17017 January 2020 Application to Implement the Full Spectrum LOCA (Fsloca) Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology CNL-20-004, Units 1 and 2, License Amendment Request to Revise Technical Specification 3.3.5, LOP DG Start Instrumentation, (WBN-TS-20-01)2020-01-17017 January 2020 Units 1 and 2, License Amendment Request to Revise Technical Specification 3.3.5, LOP DG Start Instrumentation, (WBN-TS-20-01) CNL-19-110, Application to Revise Technical Specifications 3.8.1, AC Sources - Operating (WBN-TS-18-10)2019-12-17017 December 2019 Application to Revise Technical Specifications 3.8.1, AC Sources - Operating (WBN-TS-18-10) CNL-19-077, Application to Modify Watts Bar Nuclear Plant (WBN) Unit 1 and Unit 2 Technical Specifications 3.6.15, Shield Building, (WBN-TS-19-10)2019-12-0606 December 2019 Application to Modify Watts Bar Nuclear Plant (WBN) Unit 1 and Unit 2 Technical Specifications 3.6.15, Shield Building, (WBN-TS-19-10) CNL-19-067, Application to Revise Watts Bar Nuclear Plant (WBN) Unit 2 - Technical Specifications for Steam Generator Tube Repair Sleeve (WBN-TS-391-19-13)2019-09-30030 September 2019 Application to Revise Watts Bar Nuclear Plant (WBN) Unit 2 - Technical Specifications for Steam Generator Tube Repair Sleeve (WBN-TS-391-19-13) CNL-19-060, Supplement to Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (WBN-TS-18-14)2019-08-29029 August 2019 Supplement to Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (WBN-TS-18-14) CNL-19-038, License Amendment Request to Make Miscellaneous Administrative Changes (WBN-TS-19-02)2019-06-0707 June 2019 License Amendment Request to Make Miscellaneous Administrative Changes (WBN-TS-19-02) CNL-19-014, Application to Modify Watts Bar Nuclear Plant Unit 2 Technical Specifications 3.7.8 to Extend the Completion Time for an Inoperable Essential Raw Cooling Water Train on a One-Time Basis (WBN-TS-18-07)2019-02-0707 February 2019 Application to Modify Watts Bar Nuclear Plant Unit 2 Technical Specifications 3.7.8 to Extend the Completion Time for an Inoperable Essential Raw Cooling Water Train on a One-Time Basis (WBN-TS-18-07) CNL-18-118, Application to Revise Technical Specifications Regarding DC Electrical Systems TStf-500, Revision 2, DC Electrical Rewrite - Update to TSTF-360 (WBN-TS-18-09)2018-11-29029 November 2018 Application to Revise Technical Specifications Regarding DC Electrical Systems TStf-500, Revision 2, DC Electrical Rewrite - Update to TSTF-360 (WBN-TS-18-09) CNL-18-068, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24)2018-11-29029 November 2018 Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (WBN-TS-17-24) CNL-18-119, Application to Revise Watt Bar Nuclear Plant, Units 1 & 2, Technical Specifications 3.8.1, 3.8.7, 3.8.8, and 3.8.9, Regarding Electrical Power Systems (WBN-TS-1808)2018-11-26026 November 2018 Application to Revise Watt Bar Nuclear Plant, Units 1 & 2, Technical Specifications 3.8.1, 3.8.7, 3.8.8, and 3.8.9, Regarding Electrical Power Systems (WBN-TS-1808) CNL-18-130, Revised Application to Modify the Technical Specifications for Unbalanced Voltage Relays2018-11-19019 November 2018 Revised Application to Modify the Technical Specifications for Unbalanced Voltage Relays 2023-09-20
[Table view] Category:Technical Specification
MONTHYEARCNL-23-036, Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08)2023-12-18018 December 2023 Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08) WBL-23-052, Periodic Submission for Changes Made to the Technical Specification Bases and Technical Requirements Manual2023-11-0808 November 2023 Periodic Submission for Changes Made to the Technical Specification Bases and Technical Requirements Manual CNL-23-059, Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-2023-09-20020 September 2023 Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 CNL-23-028, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06)2023-08-0202 August 2023 Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06) CNL-23-020, Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06)2023-06-28028 June 2023 Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06) ML23122A2322023-06-0707 June 2023 Issuance of Amendment Nos. 162 and 69 Regarding Change to Date in Footnotes for Technical Specification 3.7.11, Control Room Emergency Air Temperature Control System (Creatcs) CNL-23-002, Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Sche2023-03-20020 March 2023 Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Schedu CNL-22-008, and Watts Bar Nuclear Plant - Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules (BFN-TS-541, SQN-TS-21-09, and WBN-TS-22-002)2022-06-13013 June 2022 and Watts Bar Nuclear Plant - Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules (BFN-TS-541, SQN-TS-21-09, and WBN-TS-22-002) WBL-22-026, Periodic Submission for Changes Made to the Technical Specification Bases and Technical Requirements Manual2022-05-11011 May 2022 Periodic Submission for Changes Made to the Technical Specification Bases and Technical Requirements Manual ML20273A0432020-09-29029 September 2020 Plants Unit 1 and 2 - Periodic Submission for Changes Made to the Technical Specification Bases and Technical Requirements Manual CNL-20-008, Application to Modify Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specification 5.9.6, Reactor Coolant System Pressure and Temperature Limits Report, (WBN-TS-19-28)2020-07-27027 July 2020 Application to Modify Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specification 5.9.6, Reactor Coolant System Pressure and Temperature Limits Report, (WBN-TS-19-28) CNL-20-016, Application to Modify Watts Bar Nuclear Plant Unit 1 Technical Specifications 3.3.3, Post Accident Monitoring Instrumentation, (WBN-TS-19-23)2020-06-22022 June 2020 Application to Modify Watts Bar Nuclear Plant Unit 1 Technical Specifications 3.3.3, Post Accident Monitoring Instrumentation, (WBN-TS-19-23) CNL-20-004, Units 1 and 2, License Amendment Request to Revise Technical Specification 3.3.5, LOP DG Start Instrumentation, (WBN-TS-20-01)2020-01-17017 January 2020 Units 1 and 2, License Amendment Request to Revise Technical Specification 3.3.5, LOP DG Start Instrumentation, (WBN-TS-20-01) CNL-19-051, Application to Implement the Full Spectrum LOCA (Fsloca) Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology2020-01-17017 January 2020 Application to Implement the Full Spectrum LOCA (Fsloca) Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology CNL-19-097, Non-Voluntary License Amendment Request to Correct Unbalanced Voltage Relay Instrumentation Values (TS-19-22)2019-10-23023 October 2019 Non-Voluntary License Amendment Request to Correct Unbalanced Voltage Relay Instrumentation Values (TS-19-22) CNL-19-105, Supplemental Response to Second-Round NRC Request for Additional Information Regarding Application to Revise Technical Specifications Regarding DC Electrical Systems TSTF-500, Revision 2 (WBN-TS-18-09)2019-10-10010 October 2019 Supplemental Response to Second-Round NRC Request for Additional Information Regarding Application to Revise Technical Specifications Regarding DC Electrical Systems TSTF-500, Revision 2 (WBN-TS-18-09) L-19-023, Periodic Submission for Changes Made to the WBN Technical Specification Bases and Technical Requirements Manual2019-03-19019 March 2019 Periodic Submission for Changes Made to the WBN Technical Specification Bases and Technical Requirements Manual CNL-18-130, Revised Application to Modify the Technical Specifications for Unbalanced Voltage Relays2018-11-19019 November 2018 Revised Application to Modify the Technical Specifications for Unbalanced Voltage Relays CNL-18-006, Technical Specification Change - Reactor Coolant Temperature Indicator Inoperable - Exigent Amendment (391-WBN-TS-2018-01)2018-01-10010 January 2018 Technical Specification Change - Reactor Coolant Temperature Indicator Inoperable - Exigent Amendment (391-WBN-TS-2018-01) CNL-17-149, Revised Additional Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025)2017-12-27027 December 2017 Revised Additional Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025) ML17321A0332017-11-16016 November 2017 Additional Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025) ML17306A8022017-11-0202 November 2017 Periodic Submission for Changes Made to the WBN Technical Specification Bases and Technical Requirements Manual NL-17-128, Application to Modify the Watts Bar Nuclear Plant Unit 2 Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure (391-WBN-TS-17-23)2017-10-11011 October 2017 Application to Modify the Watts Bar Nuclear Plant Unit 2 Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure (391-WBN-TS-17-23) CNL-17-128, Application to Modify the Watts Bar Nuclear Plant Unit 2 Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure (391-WBN-TS-17-23)2017-10-11011 October 2017 Application to Modify the Watts Bar Nuclear Plant Unit 2 Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure (391-WBN-TS-17-23) CNL-17-123, Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025)2017-09-29029 September 2017 Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025) CNL-17-044, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.7.12, Auxiliary Building Gas Treatment System (Abgts), for Watts Bar Nuclear Plant, Units 1 and 22017-05-0505 May 2017 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.7.12, Auxiliary Building Gas Treatment System (Abgts), for Watts Bar Nuclear Plant, Units 1 and 2 CNL-17-022, Application to Modify Technical Specification 5.7.2.19 Regarding One-Time Extension of 10 CFR 50, Appendix J Type C Local Leakage Rate Tests (391-WBN-TS-17-05)2017-02-16016 February 2017 Application to Modify Technical Specification 5.7.2.19 Regarding One-Time Extension of 10 CFR 50, Appendix J Type C Local Leakage Rate Tests (391-WBN-TS-17-05) ML16113A0772016-04-22022 April 2016 Periodic Submission for Changes Made to the WBN Technical Specification Bases and Technical Requirements Manual CNL-15-060, Technical Specifications Change No. WBN2-TS-15-16 - Revise Technical Specifications for Use of Steam Generator Alternate Repair Criterion F2015-12-15015 December 2015 Technical Specifications Change No. WBN2-TS-15-16 - Revise Technical Specifications for Use of Steam Generator Alternate Repair Criterion F ML15267A1832015-09-23023 September 2015 Watts Bar, Unit 2 - Submittal of Final Revision 0 of the Technical Specifications & Technical Specification Bases, and Final Revision 0 of the Technical Requirements Manual & Technical Requirements Manual Bases NL-15-181, Watts Bar Unit 1, Revision to Essential Raw Cooling Water and Component Cooling System License Amendment Request Including Proposed Changes to Auxiliary Feedwater Pump Suction Transfer Instrumentation and Reactor Coolant System Loops - Mod2015-09-0303 September 2015 Watts Bar Unit 1, Revision to Essential Raw Cooling Water and Component Cooling System License Amendment Request Including Proposed Changes to Auxiliary Feedwater Pump Suction Transfer Instrumentation and Reactor Coolant System Loops - Mode CNL-15-181, Revision to Essential Raw Cooling Water and Component Cooling System License Amendment Request Including Proposed Changes to Auxiliary Feedwater Pump Suction Transfer Instrumentation and Reactor Coolant System Loops - Mode 42015-09-0303 September 2015 Revision to Essential Raw Cooling Water and Component Cooling System License Amendment Request Including Proposed Changes to Auxiliary Feedwater Pump Suction Transfer Instrumentation and Reactor Coolant System Loops - Mode 4 CNL-15-133, Response to Request for Additional Information Related to Technical Specification 3.8.1 Regarding Diesel Generator Steady State Frequency (WBN-TS-13-08)2015-07-15015 July 2015 Response to Request for Additional Information Related to Technical Specification 3.8.1 Regarding Diesel Generator Steady State Frequency (WBN-TS-13-08) CNL-15-088, Application to Revise Technical Specifications for Component Cooling Water Snd Essential Raw Cooling Water to Support Dual Unit Operation (TS-WBN-15-13)2015-06-17017 June 2015 Application to Revise Technical Specifications for Component Cooling Water Snd Essential Raw Cooling Water to Support Dual Unit Operation (TS-WBN-15-13) CNL-14-218, Application to Modify Technical Specification 3.8.1 Regarding Diesel Generator Steady State Frequency (WBN-TS-13-08)2015-04-0606 April 2015 Application to Modify Technical Specification 3.8.1 Regarding Diesel Generator Steady State Frequency (WBN-TS-13-08) CNL-15-001, Application to Revise Technical Specification 4.2.1, Fuel Assemblies, (WBN-TS-15-03)2015-03-31031 March 2015 Application to Revise Technical Specification 4.2.1, Fuel Assemblies, (WBN-TS-15-03) CNL-14-217, Responses to Requests for Additional Information - Developmental Revision I Technical Specification Sections 3.8 and 5.72015-03-0505 March 2015 Responses to Requests for Additional Information - Developmental Revision I Technical Specification Sections 3.8 and 5.7 CNL-14-180, Technical Specification Sections 3.0 and 3.10.12015-01-22022 January 2015 Technical Specification Sections 3.0 and 3.10.1 ML14307A9812014-11-0303 November 2014 Periodic Submission for Changes Made to Technical Specification Bases and Technical Requirement Manual ML14170A0442014-06-16016 June 2014 Attachment 5 - WBN Unit 2 TS and Tsb Through Developmental Revision I, and Trm and Trmb Through Developmental Revision D ML14170A0452014-06-16016 June 2014 Attachment 1 - TS Developmental Revision I Changes (Mark-Up) for WBN Unit 2 ML14170A0462014-06-16016 June 2014 Attachment 2 - TS Bases Developmental Revision I Changes (Mark-Up) for WBN Unit 2 ML13357A0512013-12-12012 December 2013 Synopsis of Revisions Included in Watts Bar Unit 2 Developmental Revision H of Technical Specifications and Technical Specification Bases. Part 2 of 2 ML13357A0542013-12-12012 December 2013 Technical Specifications Bases ML13353A4782013-12-12012 December 2013 Fuel Handling Accident Dose Analysis Final Safety Analysis Report and Technical Specification Revision ML12333A2422012-11-20020 November 2012 Application to Modify Technical Specification 4.3.1, Criticality, (WBN-TS-12-03) ML12333A2402012-11-19019 November 2012 Unit I, Application to Modify Technical Specification 3.7.10, Control Room Emergency Ventilation System (WBN-TS-12-01) ML12240A2242012-08-23023 August 2012 Update to Technical Requirements Manual (TRM) Bases Section 3.7.2, Developmental Revision B ML12065A0352012-02-28028 February 2012 Watts Bar, Unit 2 - Technical Specifications Table of Contents Through Page B 3.9-32 ML12065A0362012-02-28028 February 2012 Submittal of Developmental Revision G of the Unit 2 Technical Specification (TS) and Technical Specification Bases 2023-09-20
[Table view] Category:Amendment
MONTHYEARCNL-23-036, Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08)2023-12-18018 December 2023 Application to Revise Function 5 of Technical Specification Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation, for the Sequoyah Nuclear Plant and Watts Bar Nuclear Plant (SQN-TS-23-02 and WBN-TS-23-08) CNL-23-059, Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-2023-09-20020 September 2023 Supplement to Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 CNL-23-028, Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06)2023-08-0202 August 2023 Application to Revise Technical Specifications to Adopt TSTF-567-A, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (SQN-TS-23-03 and WBN-TS-23-06) CNL-23-020, Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06)2023-06-28028 June 2023 Application to Revise Technical Specifications to Adopt TSTF-501-A, Revision 1, Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee Control (WBN-TS-22-06) ML23122A2322023-06-0707 June 2023 Issuance of Amendment Nos. 162 and 69 Regarding Change to Date in Footnotes for Technical Specification 3.7.11, Control Room Emergency Air Temperature Control System (Creatcs) CNL-23-002, Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Sche2023-03-20020 March 2023 Application to Revise Watts Bar Nuclear Plant Units 1 and 2 Technical Specifications to Change the Number of Tritium Producing Burnable Absorber Rods (WBN-TS-21-02) and Proposed Revision to Reactor Vessel Surveillance Capsule Removal Schedu CNL-22-008, and Watts Bar Nuclear Plant - Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules (BFN-TS-541, SQN-TS-21-09, and WBN-TS-22-002)2022-06-13013 June 2022 and Watts Bar Nuclear Plant - Application to Revise Technical Specifications to Adopt TSTF-529, Clarify Use and Application Rules (BFN-TS-541, SQN-TS-21-09, and WBN-TS-22-002) CNL-20-008, Application to Modify Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specification 5.9.6, Reactor Coolant System Pressure and Temperature Limits Report, (WBN-TS-19-28)2020-07-27027 July 2020 Application to Modify Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specification 5.9.6, Reactor Coolant System Pressure and Temperature Limits Report, (WBN-TS-19-28) CNL-20-016, Application to Modify Watts Bar Nuclear Plant Unit 1 Technical Specifications 3.3.3, Post Accident Monitoring Instrumentation, (WBN-TS-19-23)2020-06-22022 June 2020 Application to Modify Watts Bar Nuclear Plant Unit 1 Technical Specifications 3.3.3, Post Accident Monitoring Instrumentation, (WBN-TS-19-23) CNL-20-004, Units 1 and 2, License Amendment Request to Revise Technical Specification 3.3.5, LOP DG Start Instrumentation, (WBN-TS-20-01)2020-01-17017 January 2020 Units 1 and 2, License Amendment Request to Revise Technical Specification 3.3.5, LOP DG Start Instrumentation, (WBN-TS-20-01) CNL-19-097, Non-Voluntary License Amendment Request to Correct Unbalanced Voltage Relay Instrumentation Values (TS-19-22)2019-10-23023 October 2019 Non-Voluntary License Amendment Request to Correct Unbalanced Voltage Relay Instrumentation Values (TS-19-22) CNL-19-105, Supplemental Response to Second-Round NRC Request for Additional Information Regarding Application to Revise Technical Specifications Regarding DC Electrical Systems TSTF-500, Revision 2 (WBN-TS-18-09)2019-10-10010 October 2019 Supplemental Response to Second-Round NRC Request for Additional Information Regarding Application to Revise Technical Specifications Regarding DC Electrical Systems TSTF-500, Revision 2 (WBN-TS-18-09) CNL-17-149, Revised Additional Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025)2017-12-27027 December 2017 Revised Additional Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025) ML17321A0332017-11-16016 November 2017 Additional Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025) NL-17-128, Application to Modify the Watts Bar Nuclear Plant Unit 2 Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure (391-WBN-TS-17-23)2017-10-11011 October 2017 Application to Modify the Watts Bar Nuclear Plant Unit 2 Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure (391-WBN-TS-17-23) CNL-17-128, Application to Modify the Watts Bar Nuclear Plant Unit 2 Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure (391-WBN-TS-17-23)2017-10-11011 October 2017 Application to Modify the Watts Bar Nuclear Plant Unit 2 Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Fluid Oil Pressure (391-WBN-TS-17-23) CNL-17-123, Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025)2017-09-29029 September 2017 Supplement to Application to Revise Technical Specifications to Adopt TSTF-547, Revision 1, Clarification of Rod Position Requirements (WBN-TS-16-025) CNL-17-044, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.7.12, Auxiliary Building Gas Treatment System (Abgts), for Watts Bar Nuclear Plant, Units 1 and 22017-05-0505 May 2017 Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification 3.7.12, Auxiliary Building Gas Treatment System (Abgts), for Watts Bar Nuclear Plant, Units 1 and 2 CNL-15-133, Response to Request for Additional Information Related to Technical Specification 3.8.1 Regarding Diesel Generator Steady State Frequency (WBN-TS-13-08)2015-07-15015 July 2015 Response to Request for Additional Information Related to Technical Specification 3.8.1 Regarding Diesel Generator Steady State Frequency (WBN-TS-13-08) CNL-15-088, Application to Revise Technical Specifications for Component Cooling Water Snd Essential Raw Cooling Water to Support Dual Unit Operation (TS-WBN-15-13)2015-06-17017 June 2015 Application to Revise Technical Specifications for Component Cooling Water Snd Essential Raw Cooling Water to Support Dual Unit Operation (TS-WBN-15-13) CNL-14-218, Application to Modify Technical Specification 3.8.1 Regarding Diesel Generator Steady State Frequency (WBN-TS-13-08)2015-04-0606 April 2015 Application to Modify Technical Specification 3.8.1 Regarding Diesel Generator Steady State Frequency (WBN-TS-13-08) CNL-15-001, Application to Revise Technical Specification 4.2.1, Fuel Assemblies, (WBN-TS-15-03)2015-03-31031 March 2015 Application to Revise Technical Specification 4.2.1, Fuel Assemblies, (WBN-TS-15-03) ML14170A0462014-06-16016 June 2014 Attachment 2 - TS Bases Developmental Revision I Changes (Mark-Up) for WBN Unit 2 ML14170A0452014-06-16016 June 2014 Attachment 1 - TS Developmental Revision I Changes (Mark-Up) for WBN Unit 2 ML14170A0442014-06-16016 June 2014 Attachment 5 - WBN Unit 2 TS and Tsb Through Developmental Revision I, and Trm and Trmb Through Developmental Revision D ML11228A0472011-08-10010 August 2011 Technical Specifications (Developmental Revision F), Enclosure 1 ML1022900752010-08-16016 August 2010 Unit 2 - Change to Developmental TS Section 4.2.2, Control Rod Assemblies ML1005504412010-02-0202 February 2010 Developmental Revision B - Technical Specifications 2.0 - Safety Limits ML1005503672010-02-0202 February 2010 Developmental Revision B - Technical Specifications Table of Contents ML1005504742010-02-0202 February 2010 Developmental Revision B - Technical Specifications 5.0 - Administrative Controls ML1005504712010-02-0202 February 2010 Developmental Revision B - Technical Specifications 4.0 - Design Features ML1005504012010-02-0202 February 2010 Developmental Revision B - Technical Specifications 1.0 - Use and Application ML1005504472010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.0 - Limiting Condition for Operation (LCO) Applicability ML1005504642010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.9 - Refueling Operations ML1005504492010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.1 - Reactivity Control Systems ML1005504622010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.8 - Electrical Power Systems ML1005504612010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.7 - Plant Systems ML1005504502010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.2 - Power Distribution Limits ML1005504512010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.3 - Instrumentation ML1005504532010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.4 - Reactor Coolant System ML1005504562010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.5 - Emergency Core Cooling Systems ML1005504582010-02-0202 February 2010 Developmental Revision B - Technical Specifications 3.6 - Containment Systems ML0929503402009-10-20020 October 2009 License Amendment Request for Adoption of TSTF-511, Revision 0, Eliminate Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26 - Browns Ferry TS Change 469; Sequoyah Change 09-04; and Watts Change 09-19 ML0913205622009-04-30030 April 2009 Application for Technical Specification Change to Correct Minor Error in Amendment 70 ML0900900442008-12-31031 December 2008 Revised Technical Specifications Change WBN-TS-08-04 - Revision to the Maximum Number of TPBARS That Can Be Irradiated in the Reactor Core Per Cycle ML0831708612008-11-12012 November 2008 License Amendment Request TS-08-11 to Revise Reactor Coolant System Leakage Detection Systems - Exigent Change Request ML0712106042007-04-25025 April 2007 (WBN) Unit 1 - Technical Specification Change 07-01, Revision of Number of Tritium Producing Burnable Absorber Rods (TPBARS) in the Reactor Core ML0616502672006-06-12012 June 2006 Technical Specifications (TS) Change WBN-TS-05-10 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity - Request for Additional Information ML0526500612005-09-20020 September 2005 T.S. for License Amendment, Regarding Alternative Means for Monitoring Control or Shutdown Rod Position ML0509802082005-03-21021 March 2005 Tech Spec Pages for Amendment No. 57, Elimination of Monthly Report Requirement on Operating & Occupational Reports (TS-04-15) 2023-09-20
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Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 JUN 12 2006 WBN-TS-05-10 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of ) Docket No. 50-390 Tennessee Valley Authority WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - TECHNICAL SPECIFICATIONS (TS) CHANGE WBN-TS-05 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY - REQUEST FOR ADDITIONAL INFORMATION (TAC NO. MC 9271)
The purpose of this letter is to provide TVA's response to the request for additional information dated May 9, 2006, concerning the subject amendment request that was submitted to NRC on December 15, 2005. provides TVA's response to NRC's questions. There are no regulatory commitments associated with this submittal.
If you have any questions concerning this matter, please call me at (423) 365-1824.
Printed on recycled paper
U.S. Nuclear Regulatory Commission Page 2 JUN 12 2006 I declare under penalty of perjury that the foregoing is true and correct. Executed on this 1 2 th day of June 2006.
Sincerely, P. L. Pace Manager, Site Licensing and Industry Affairs Enclosures
- 1. Response to RAI Questions
- 2. Revised Technical Specification Page
- 3. Revised Technical Specification Bases Pages cc (Enclosures):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 TVA submitted an application for an amendment to revise the WBN Unit 1 technical specification (TS) requirements to be consistent with Technical Specification Task Force (TSTF) Traveler, TSTF-449, Revision 4, "Steam Generator Tube Integrity," by letter dated December 15, 2005. As stated on page E1-3 of Enclosure 1 to the application, the current steam generators will be replaced in the Fall 2006. In addition, the approved alternate repair criteria (ARC) (i.e., voltage based ARC for outside diameter stress corrosion cracking and the use of the F-star), and the sleeving repair method will be deleted as part of this TS change.
NRC issued a request for additional information concerning the subject TS change dated May 9, 2006. TVA's response is provided below:
NRC QUESTION 1 Insert A of Enclosure 2 to the application contains TS 5.7.2.12, "Steam Generator (SG) Program," and corresponds to Insert 5.5.9 of TSTF-449, Revision 4.
The last sentence in TS 5.7.2.12.a states, "Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected and/or plugged, to confirm that the performance criteria are being met." The intent of this paragraph is to ensure that condition monitoring assessments are conducted when the SG tubes are inspected or plugged as stated in paragraph a of Insert 5.5.9 of TSTF-449, Revision 4. The staff requests the licensee to either justify the use of "and/or" in the last sentence of TS 5.7.2.12.a or to replace "and/or" with "or.
RESPONSE
See attached change to TS 5.7.2.12.a. - "and/or" has been replaced with "or."
NRC QUESTION 2 Insert A of Enclosure 2 to the application contains TS 5.7.2.12, "Steam Generator (SG) Program," and corresponds to Insert 5.5.9 of TSTF-449, Revision 4.
The last sentence in TS 5.7.2.12.b.2 states, "The accident induced leakage is not to exceed 1.0 gpm for the faulted SG. "
The corresponding sentence in Insert 5.5.9 of TSTF-449, Revision EI-l
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 4, states, "Leakage is not to exceed [1 gpm] per SG [, except for specific types of degradation at specific locations described in paragraph c of the Steam Generator Program]." The intent of this sentence is to ensure that leakage does not exceed 1 gpm in any SG except for those instances defined in paragraph c of the SG program. The staff requests the licensee to discuss its plans to modify this sentence consistent with TSTF-449, Revision 4.
RESPONSE
TVA acknowledges that the TSTF, Insert 5.5.9 states "per SG."
However, the use of the words "per SG" indicates WBN could have a total from all four steam generators of four gallons per minute (gpm) primary-to-secondary leakage during an Main Steam Line Break (MSLB) accident. The MSLB dose calculation for the replacement steam generators uses one gpm in the faulted steam generator and 150 gallons per day (gpd) which is 0.10 gpm in the non-faulted steam generators (See Technical Specification Bases 3.4.13, RCS Operational Leakage, under Applicable Safety Analysis, page B 3.4-75, contained in TVA's request, WBN-TS-05-10, dated December 15, 2005). Therefore, TVA considers the wording change justified to prevent a misinterpretation which would be outside the WBN design basis.
NRC QUESTION 3 to the application includes the new bases for TS 3.4.17, "Steam Generator (SG) Tube Integrity," and corresponds to section B 3.4.20, "Steam Generator (SG) Tube Integrity," found on pages B 3.4.20-1 through B 3.4.20-7 of TSTF-449, Revision 4.
On page B 3.4-100 of the application, second paragraph under "Applicable Safety Analysis," the last sentence reads, "The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3)." The corresponding sentence on page B 3.4.20-2 of TSTF-449, Revision 4, includes, "or the NRC approved licensing basis (e.g., a small fraction of these limits)." The staff requests the licensee to justify the exclusion of this phrase from its application.
RESPONSE
The calculated Main Control Room thyroid dose for the replacement steam generators in the MSLB radiological dose calculation is 12.5 rem versus the limit of 30 rem. Therefore, TVA did not include the words in parenthesis that state "(e.g., a small EI-2
ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 fraction of these limits)," as 12.5 rem can not be considered a small fraction of 30 rem. However, 12.5 rem is well within the NRC's regulatory limits. TVA has added the words to the end of Bases second paragraph under Applicable Safety Analyses, page B 3.4-100, "or the NRC approved licensing basis." See attached mark-up of page B 3.4-100.
NRC QUESTION 4 to the application includes the new bases for TS 3.4.17, "Steam Generator (SG) Tube Integrity," and corresponds to section B 3.4.20, "Steam Generator (SG) Tube Integrity," found on pages B 3.4.20-1 through B 3.4.20-7 of TSTF-449, Revision 4.
On page B 3.4-102 of the application, second paragraph under "Actions," the last sentence reads, "If it is determined that tube integrity is not being maintained until the next SG inspection, Condition B applies." The corresponding sentence on page B.3.4.20-4 of TSTF-449, Revision 4, does not include the phrase "until the next SG inspection," however, the required action found in Enclosure 2 of the application, page 3.4-43, includes the phrase, "until the next refueling outage or SG tube inspection." The staff requests the licensee to justify the deviation from the action described in TS 3.4.17 and the wording in TSTF-449, Revision 4.
RESPONSE
See attached change to B 3.4-102. "..until the next SG inspection," has been deleted to be consistent with the TSTF.
EI-3
ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 REVISED PROPOSED TECHNICAL SPECIFICATION PAGES (MARK UP)
I. Affected Page List 5.0-15 Insert A II. Marked Pages See Attached
INSERT A 5.7 Procedures, Programs, and Manuals (continued) 5.7.2.12 Steam Generator (SG) Proaram A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
- a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.
- b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, cooldown, and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.
Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- 2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. The accident induced leakage is not to exceed 1.0 gpm for the faulted SG.
- 3. The operational leakage performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
- c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
- d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and
ENCLOSURE 3 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST WBN-TS-05-10 STEAM GENERATOR TUBE INTEGRITY - TSTF-449, REVISION 4 REVISED PROPOSED TECHNICAL SPECIFICATION PAGES I. Affected Page List B 3.4-100 B 3.4-102 II. Marked Pages See Attached
SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY ANALYSES basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.
The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from 150 gallons per day (gpd) per steam generator and 1 gallon per minute (gpm) in the faulted steam generator. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16 "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3) or the NRC approved licensing basis.
Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.
During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.
A SG tube has tube integrity when it satisfies the SG performance criteria.
The SG performance criteria are defined in Specification 5.7.2.12, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.
(continued)
Watts Bar-Unit 1 B 3.4-100 Revision Amendment
SG Tube Integrity B 3.4.17 BASES LCO The operational LEAKAGE performance criterion provides an observable (continued) indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day.
This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.
APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.
RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.
ACTIONS The ACTIONS are modified by a Note that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube.
Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry, and application of associated Required Actions.
A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged, has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained until the next SG
..... ie~I,
-Condition B applies. I (continued)
Watts Bar-Unit I B 3.4-102 Revision Amendment