SBK-L-22054, SBK-L-22054 - License Amendment Request 22-01, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections
| ML22160A581 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 06/09/2022 |
| From: | Lesniak T NextEra Energy Seabrook |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SBK-L-22054 | |
| Download: ML22160A581 (21) | |
Text
NEXTera*
EN~
Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington DC 20555-0001 RE:
Seabrook Station Docket No. 50-443 Renewed Facility Operating License No. NPF-86 June 9, 2022 SBK-L-22054 10 CFR 50.90 License Amendment Request 22-01, Application to Revise Technical Specifications to Adopt TSTF-577, "Revised Frequencies for Steam Generator Tube Inspections" Pursuant to 10 CFR 50.90, NextEra Energy Seabrook, LLC (NextEra) is submitting a request for an amendment to the Technical Specifications (TS) for Seabrook Station Unit 1 (Seabrook).
NextEra requests adoption of TSTF 577, "Revised Frequencies for Steam Generator Tube Inspections,"
which is an approved change to the Standard Technical Specifications (STS), into the Seabrook, Unit 1 TS.
The TS related to steam generator (SG) tube inspections and reporting are revised based on operating history.
The enclosure provides a description and assessment of the proposed changes. Attachment 1 to the enclosure provides the existing TS pages marked to show the proposed changes. Attachment 2 provides revised (clean copy) TS pages. The TS Bases are not affected by the proposed changes.
NextEra requests that the amendment be reviewed under the Consolidated Line-Item Improvement Process (CLIIP). Approval of the proposed amendment is requested within 6 months of completion of the NRC's acceptance review. Once approved, the amendment shall be implemented within 60 days.
There are no regulatory commitments made in this submittal.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated New Hampshire official.
If you should have any questions regarding this submittal, please contact Mr. Michael Davis, Fleet Licensing Manager, at 319-851-7032.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on the 9th day of June 2022.
J2~£r 1/2imothy Lesniak General Manager, Regulatory Affairs
Enclosure:
Description and Assessment Attachments: 1. Proposed Technical Specification Changes (Mark-Up)
- 2. Revised Technical Specification Pages NextEra Energy Seabrook, LLC P.O. Box 300, Lafayette Road, Seabrook, NH 03874
Seabrook Station Docket Nos. 50-443 cc:
USNRC Region I Administrator USNRC Project Manager USNRC Senior Resident Inspector Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 John Viveiros, Senior Nuclear Planner The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 SBK-L-22054 Page 2 of 2
Seabrook Station Docket Nos. 50-443 Description and Assessment Seabrook Station SBK-L-22054 Enclosure Page 1 of 19 License Amendment Request 22-01, Application to Revise Technical Specifications to Adopt TSTF 577, "Revised Frequencies for Steam Generator Tube Inspections"
1.0 DESCRIPTION
................................................................................................................................. 2
2.0 ASSESSMENT
................................................................................................................................ 2 2.1 Applicability of Safety Evaluation......................................................................................... 2 2.2 Variations............................................................................................................................. 2
3.0 REGULATORY ANALYSIS
............................................................................................................. 3 3.1 No Significant Hazards Consideration Analysis.................................................................. 3 3.2 Conclusion........................................................................................................................... 4
4.0 ENVIRONMENTAL CONSIDERATION
.......................................................................................... 4 ATTACHMENTS
- 1.
Proposed Technical Specification Changes (Mark-Up)
- 2.
Revised Technical Specification Pages
Seabrook Station Docket Nos. 50-443
1.0 DESCRIPTION
SBK-L-22054 Enclosure Page 2 of 19 NextEra Energy Seabrook, LLC (NextEra) requests adoption of TSTF-577, "Revised Frequencies for Steam Generator Tube Inspections," which is an approved change to the Standard Technical Specifications (STS), into the Seabrook Nuclear Plant Unit 1 (Seabrook)
Technical Specifications (TS). The TS related to steam generator (SG) tube inspections and reporting are revised based on operating history.
2.0 ASSESSMENT
2.1 Applicability of Safety Evaluation NextEra has reviewed the safety evaluation for TSTF-577 provided to the Technical Specifications Task Force in a letter dated April 14, 2021. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-577.
As described herein, NextEra has concluded that the justifications presented in TSTF-577 and the safety evaluation prepared by the NRC staff are applicable to Seabrook Unit 1 and justify this amendment for the incorporation of the changes to the Seabrook TS.
The current SG TS requirements are based on TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
The Seabrook SGs have Alloy 600 thermally treated (Alloy 600TT) tubes.
The initial inspection period described in the SG Program, paragraph d.2, began in October 2021. NextEra will submit a SG Tube Inspection Report meeting the revised TS 6.8.1.7 requirements within 30 days after implementation of the license amendment.
2.2 Variations NextEra is proposing the following variations from the TS changes described in TSTF-577 or the applicable parts of the N RC staffs safety evaluation:
- 1. The Seabrook TS utilize different numbering than the Standard Technical Specifications on which TSTF-577 was based. Specifically, the Steam Generator (SG) Program is Specification 6.7.6.k rather than 5.5.9, and the Steam Generator Tube Inspection Report is Specification 6.8.1.7 rather than 5.6.7. In addition, in Specification 6.7.6.k.b.3, the operational LEAKAGE performance criterion is specified in LCO 3.4.6.2 rather than 3.4.13.
- 2.
In Specification 6.7.6.k.b.3, the phrase "Reactor Coolant System" is used in place of the acronym "RCS".
- 3. Seabrook Specification 6.8.1.7 is revised to place the title of Specification 6.7.6.k in quotes. This change is consistent with TSTF-577.
These differences are administrative and do not affect the applicability of TSTF-577 to the Seabrook TS.
The Seabrook SG Program TS currently contain a provision for an alternate tube plugging criteria. The description of the alternate tube plugging criteria in the proposed change is equivalent to the descriptions in the current TS.
Seabrook Station Docket Nos. 50-443 SBK-L-22054 Enclosure Page 3 of 19 3.0 The Seabrook TS contain requirements that differ from the Standard Technical Specifications on which TSTF-577 was based but are encompassed in the TSTF-577 justification. Specification 6.7.6.k.b.2 contains limits on leakage rate for any one SG as well as the total for all SGs - rather than a leakage rate per SG as used in the Standard Technical Specifications. Specifically, the last sentence of Specification 6.7.6.k.b.2 states, "Leakage is not to exceed 1 gpm total or 500 gpd through any one SG" rather than "Leakage is not to exceed 1 gpm per SG." The Seabrook TS requirement for accident induced leakage rate is more conservative and therefore acceptable. A similar variation was approved for North Anna Power Station Unit Nos.
1 and 2 in Amendment Nos. 292 and 275, respectively. Reference NRC ADAMS Accession No. ML22068A071.
REGULATORY ANALYSIS 3.1 No Significant Hazards Consideration Analysis NextEra Energy Seabrook, LLC (NextEra) requests adoption of TSTF-577, "Revised Frequencies for Steam Generator Tube Inspections," which is an approved change to the Standard Technical Specifications (STS), into the Seabrook Station, Unit 1 (Seabrook) Technical Specifications (TS). The TS related to steam generator (SG) tube inspections and reporting are revised based on operating history.
NextEra has evaluated if a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
(1)
Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change revises the inspection frequencies for SG tube inspections and associated reporting requirements. The SG inspections are conducted as part of the SG Program to ensure and demonstrate that performance criteria for tube structural integrity and accident leakage integrity are met. These performance criteria are consistent with the plant design and licensing basis. With the proposed changes to the inspection frequencies, the SG Program must still demonstrate that the performance criteria are met. As a result, the probability of any accident previously evaluated is not significantly increased and the consequences of any accident previously evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2)
Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No
Seabrook Station Docket Nos. 50-443 SBK-L-22054 Enclosure Page 4 of 19 The proposed change revises the inspection frequencies for SG tube inspections and associated reporting requirements. The proposed change does not alter the design function or operation of the SGs or the ability of an SG to perform the design function. The SG tubes continue to be required to meet the SG Program performance criteria. The proposed change does not create the possibility of a new or different kind of accident due to credible new failure mechanisms, malfunctions, or accident initiators that are not considered in the design and licensing bases.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
(3)
Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed change revises the inspection frequencies for SG tube inspections and associated reporting requirements. The proposed change does not change any of the controlling values of parameters used to avoid exceeding regulatory or licensing limits. The proposed change does not affect a design basis or safety limit, or any controlling value for a parameter established in the UFSAR or the license.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, NextEra concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c),
and, accordingly, a finding of "no significant hazards consideration" is justified.
3.2 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
4.0 ENVIRONMENT AL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
Seabrook Station Docket No. 50-443 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)
(7 pages follow)
SBK-L-22054 Enclosure Page 5 of 19
Seabrook Station Docket No. 50-443 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)
Technical Specification Mark-up Inserts INSERT 1:
SBK-L-22054 Enclosure Page 6 of 19 After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 54 effective full power months, which defines the inspection period. If none of the SG tubes have ever experienced cracking other than in regions that are exempt from inspection by alternate repair criteria and the SG inspection was performed with enhanced probes, the inspection period may be extended to 72 effective full power months. Enhanced probes have a capability to detect flaws of any type equivalent to or better than array probe technology. The enhanced probes shall be used from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet except any portions of the tube that are exempt from inspection by alternate repair criteria. If there are regions where enhanced probes cannot be used, the tube inspection techniques shall be capable of detecting all forms of existing and potential degradation in that region.
INSERT 2:
- b.
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;
- c.
For each degradation mechanism found:
- 1.
The nondestructive examination techniques utilized;
- 2.
The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
- 3.
A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and
- 4.
The number of tubes plugged during the inspection outage.
- d.
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS
- 6. 7.6 (Continued)
- j. Technical Specification (TS) Bases Control Program
- k.
This program provides a means for processing changes to the Bases of these Technical Specifications.
- a.
Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
- b.
Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
- 1.
A change in the TS incorporated in the license or
- 2.
A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59.
- c.
The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
- d.
Proposed changes that meet the criteria of Specification 6.7.6j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).
Steam Generator (SG) Program An SG SteaR1 Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG SteaR1 Generator Program shall include the following:
- 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.
SEABROOK - UNIT 1 6-11 Amendment No. 34, 55, 67, 88, 104, 44-a, 138
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
- 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 SG stearTI generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), 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.
Leakage is not to exceed 1 gpm total or 500 gpd through any one SG.
- 3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."
- c.
Provisions for SG tube plugging 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.
SEABROOK - UNIT 1 6-12 Amendment No. 34, 39, 104, 109, 115, 138
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS
- 6. 7.6 (Continued)
- d.
except for any portions of the tube that are exempt from inspection by alternate repair criteria INSERT 1 The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria:
Tubes with service-induced flaws located greater than 15.21 inches below the top of the tubesheet do not require plugging.
Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.21 inches below the top of the tubesheet shall be plugged upon detection.
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 outl t, and that may satisfy the applicable tube plugging criteria.
is exoluded from this requirement 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. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
After the first refueling outage following SG installation, inspeot eaoh SG at least e'lery 48 effeoth1e full po1Ner months or at least el/eF)' other refueling outage (whiohe1,er results in more frequent inspeotions). In addition, the minimum number of tubes inspeoted at eaoh soheduled inspeotion shall be the number of tubes in all SGs di'lided by the number of SG inspeotion outages soheduled in eaoh inspeotion period as defined in a, b, and a below. If a degradation assessment indioates the potential for a type of degradation to ooour at a looation not pre'liously inspeoted with a teohnique oapable of detesting this type of degradation at this looation and that may satisfy the applioable tube plugging oriteria, the minimum number of looations inspeoted \\11ith sueh a oapable inspeotion SEABROOK - UNIT 1 6-13 Amendment No. 34,104,109,115, 123,131,138
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS
- 6. 7.6 (Continued) excluding any region that is exempt from inspection by alternate repair criteria be at the next teohniql:le dl:lring the remainder of the inspeotion period may be prorated. The fraotion of looations to be inspeoted for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the nl:lmber of times the SG is sohedl:lled to be inspeoted in the inspeotion period after the determination that a new form of degradation ool:lld potentially be oool:lrring at this looation di*,ided by the total nl:lmber of times the SG is sohedl:lled to be inspeoted in the inspection period. each inspection period defined belm\\' may be mEtended l:lp to 3 effeoti*,e fl:lll po1,¥er months to inoll:lde a SG inspeotion ol:ltage in an inspeotion period and the sl:lbseql:lent inspeotion period begins at the oonoll:lsion of the inoll:lded SG inspeotion ol:ltage.
After the first refl:leling ol:ltage following SG installation, inspect 100% of the tl:lbes dl:lring the nmEt 120 effecti¥e fl:lll power months. This oonstitl:ltes the first inspeotion period; Dl:lring the nmd 9@ effecti¥e fl:lll power months, inspeot 100% of the tl:lbes. This oonstitl:ltes the seoond inspection period; and Dl:lring the remaining life of the SGs, inspect 100 % of the tl:lbes e¥ery 72 effeoti¥e fl:lll power months. This constitl:ltes the third and sl:lbseql:lent inspection periods.
refueling outage, but
- 3.
If crack indications are found in portions of the SG tube may be deferred to not exoll:lded abo¥e, then the next inspection for each affected the following and potentially affected SG for the degradation mechanism that refueling outage if caused the crack indication shall not exoeed 24 effeoti*,e fl:lll the 100% inspection L-----"7' power months or one refl:leling ol:ltage (whiche¥er resl:llts in more of all SGs was freql:lent inspeotions). If definitive information, such as from performed with examination of a pulled tube, diagnostic non-destructive testing, enhanced probes as or engineering evaluation indicates that a crack-like indication is described in not associated with a crack(s), then the indication need not be paragraph d.2 treated as a crack.
- e.
Provisions for monitoring operational primary to secondary leakage.
SEABROOK - UNIT 1 6-14 Amendment No. 34, 78,104,115, 119,123,138
ADMINISTRATIVE CONTROLS 6.8.1.6.c The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.
STEAM GENERATOR TUBE INSPECTION REPORT 6.8.1.7
- a.
lINSERT2 ~
~
~
- f.
A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.7.6.k, "Steam Generator (SG) Program." The report shall include:
The scope of inspections performed on each SG;,
Degradation meehanisms f.ound,
Nondestrueti¥e examination teehniques utilii!:ed f.or eaeh degradation meehanism, Loeation, orientation (if linear), and measured sii!:es (if a1,ailable) of serviee indueed indieations,
~Jumber of tubes plugged during the inspeetion outage f.or eaeh degradation meehanism, The number and percentage of tubes plugged to date, and the effective plugging percentage in each SGsteam generator,;
The results of eondition monitoring, ineluding the results of tube pulls and in situ testing, The primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report;,
The calculated accident induced leakage rate from the portion of the tubes below 15.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined;, and SEABROOK - UNIT I 6-21 Amendment No. 22, 66, 88, 104, 107, 115, 123, 131, 138
---f.
The results of any SG secondary side inspections;
ADMINISTRATIVE CONTROLS 6.8.1.7 (Continued)
The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.
6.9 (THIS SPECIFICATION NUMBER IS NOT USED)
SEABROOK - UNIT I 6-21a Amendment No. 123, 131, 138
Seabrook Station Docket Nos. 50-443 ATTACHMENT 2 REVISED TECHNICAL SPECIFICATION PAGES (6 pages follow)
SBK-L-22054 Enclosure Page 13 of 19
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
- j.
Technical Specification (TS) Bases Control Program
- k.
This program provides a means for processing changes to the Bases of these Technical Specifications.
- a.
Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
- b.
Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
- 1.
A change in the TS incorporated in the license or
- 2.
A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59.
- c.
The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
- d.
Proposed changes that meet the criteria of Specification 6.7.6j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
Steam Generator (SG) Program An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:
- 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.
SEABROOK-UNIT 1 6-11 Amendment No. 34, 55, 67, 88, 104, 44-a, 138
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
- 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 SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), 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.
Leakage is not to exceed 1 gpm total or 500 gpd through any one SG.
- 3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."
- c.
Provisions for SG tube plugging 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.
The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria:
SEABROOK - UNIT 1 6-12 Amendment No. 34, 39, 104, 109, 115, 138
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
Tubes with service-induced flaws located greater than 15.21 inches below the top of the tubesheet do not require plugging.
Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.21 inches below the top of the tubesheet shall be plugged upon detection.
- 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 except for any portions of the tube that are exempt from inspection by alternate repair criteria, and that may satisfy the applicable tube plugging 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. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 1.
- 2.
SEABROOK - UNIT 1 Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 54 effective full power months, which defines the inspection period. If none of the SG tubes have ever experienced cracking other than in regions that are exempt from inspection by alternate repair criteria and the SG inspection was performed with enhanced probes, the inspection period may be extended to 72 effective full power months. Enhanced probes have a capability to detect flaws of any type equivalent to or better than array probe technology. The enhanced probes shall be used from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet except any portions of the tube that are exempt from inspection by alternate repair criteria. If there are regions where enhanced probes cannot be used, the tube inspection techniques shall be capable of detecting all forms of existing and potential degradation in that region.
6-13 Amendment No. 34,104,109,115, 123, 131, 138
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
- 3.
If crack indications are found in portions of the SG tube excluding any region that is exempt from inspection by alternate repair criteria, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage, but may be deferred to the following refueling outage if the 100% inspection of all SGs was performed with enhanced probes as described in paragraph d.2. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
- e.
Provisions for monitoring operational primary to secondary leakage.
SEABROOK - UNIT 1 6-14 Amendment No. 34, 78, 104, 115, 119,123,138
ADMINISTRATIVE CONTROLS 6.8.1.6.c The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.
STEAM GENERATOR TUBE INSPECTION REPORT 6.8.1. 7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.7.6.k, "Steam Generator (SG) Program." The report shall include:
- a.
The scope of inspections performed on each SG;
- b.
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;
- c.
For each degradation mechanism found:
- 1.
The nondestructive examination techniques utilized;
- 2.
The location, orientation (if linear), measured size (if available),
and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
- 3.
A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and
- 4.
The number of tubes plugged during the inspection outage.
- d.
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;
- e.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; SEABROOK - UNIT I 6-21 Amendment No. 22, 66, 88, 104, 107, 115,123,131,138
ADMINISTRATIVE CONTROLS 6.8.1.7
- f.
- g.
- h.
- i.
(Continued)
The results of any SG secondary side inspections; The primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report; The calculated accident induced leakage rate from the portion of the tubes below 15.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.
6.9 (THIS SPECIFICATION NUMBER IS NOT USED)
SEABROOK - UNIT I 6-21a Amendment No. 123, 131, 138