HNP-14-050, Application to Revise Technical Specifications to Adopt TSTF-510-A, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using the Consolidated Line Item Improvement Process
| ML14114A743 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 04/24/2014 |
| From: | Kapopoulos E Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| HNP-14-050 | |
| Download: ML14114A743 (27) | |
Text
Ernest J. Kapopoulos, Jr.
Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill NC 27562-9300 919.362.2000 10 CFR 50.90 Serial: HNP-14-050 April 24, 2014 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400
Subject:
Application to Revise Technical Specifications to Adopt TSTF-510-A, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process Ladies and Gentlemen:
Pursuant to 10 CFR 50.90, Duke Energy Progress, Inc. (Duke Energy), hereby requests a revision to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1.
The proposed amendment would modify TS requirements regarding steam generator tube inspections and reporting as described in Technical Specifications Task Force (TSTF) change TSTF-510-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed changes involve no significant hazards consideration. The bases for these determinations are included in the enclosure. Duke Energy is providing the state of North Carolina with a copy of this submittal in accordance with 10 CFR 50.91(b).
The enclosure provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Attachment 1 to the enclosure provides the existing TS pages marked up to show the proposed changes. Attachment 2 to the enclosure provides revised TS pages. Attachment 3 to the enclosure provides existing TS Bases pages marked up to show the proposed changes.
Approval of the proposed license amendment is requested by October 25, 2014. Once approved, the amendment shall be implemented within 120 days.
This document contains no new Regulatory Commitments.
U.S. Nuclear Regulatory Commission HNP-14-050 Page2 Please refer any questions regarding this submittal to John Caves at (919) 362-2406.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on [
].
Sincerely, Ernest J. Kapopoulos, Jr.
Enclosure:
Evaluation of the Proposed Change cc:
Mr. J. D. Austin, NRC Sr. Resident Inspector,HNP Mr. W. L. Cox, III,SectionChief, N.C. DHSR(Emaillee.cox@dhhs.nc.gov)
Mr. A. Hon, NRC Project Manager, HNP Mr. V. M. McCree, NRC Regional Administrator, Region II
HNP-14-050 Enclosure Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Evaluation of the Proposed Change Application to Revise Technical Specifications to Adopt TSTF-510-A, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process Proposed Technical Specification Changes Revised Technical Specification Pages Proposed Technical Specification Bases Changes
U.S. Nuclear Regulatory Commission Page 1 of 4 HNP-14-050 Enclosure Evaluation of the Proposed Change Application to Revise Technical Specifications to Adopt TSTF-510-A, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process 1.0 Description The proposed change revises Specification 6.8.4.l, "Steam Generator (SG) Program," 6.9.1.7, "Steam Generator Tube Inspection Report," and Technical Specification (TS) 3/4.4.5, "Steam Generator Tube Integrity." The proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications.
The proposed amendment is consistent with TSTF-510-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
2.0 Assessment 2.1. Applicability of Published Safety Evaluation Duke Energy Progress, Inc. (Duke Energy), has reviewed TSTF-510-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," and the model safety evaluation dated October 27, 2011. Duke Energy has concluded that the justifications presented in TSTF-510-A and the model safety evaluation prepared by the NRC staff are applicable to Shearon Harris Nuclear Power Plant, Unit 1 (HNP) and justify this amendment for the incorporation of the changes to the HNP TS.
2.2. Optional Changes and Variations Duke Energy is proposing the following variations from the TS changes described in the TSTF-510-A, Revision 2, or the applicable parts of the NRC staffs model safety evaluation dated October 27, 2011.
The HNP TS utilize different numbering than the Standard Technical Specifications (STS) on which TSTF-510-A was based. Specifically, the STS and corresponding HNP numbering is as follows.
STS Numbering HNP Numbering TS 3.4.20, "Steam Generator (SG) Tube Integrity" TS 3/4.4.5, "Steam Generator (SG) Tube Integrity" Limiting Condition for Operation (LCO) 3.4.20 LCO 3.4.5 LCO 3.4.20, Condition A LCO 3.4.5, Action a Surveillance Requirement (SR) 3.4.20.2 SR 4.4.5.2 Specification 5.5.9, "Steam Generator (SG)
Program" Specification 6.8.4.l, "Steam Generator (SG)
Program." Note: HNP also uses a different numbering scheme within Specification 6.8.4.l Specification 5.6.7, "Steam Generator Tube Inspection Report" Specification 6.9.1.7, "Steam Generator Tube Inspection Report"
U.S. Nuclear Regulatory Commission Page 2 of 4 HNP-14-050 Enclosure These differences are administrative and do not affect the applicability of TSTF-510-A to the HNP TS.
The proposed change corrects an administrative inconsistency in TSTF-510-A, Paragraph d.2 of the Steam Generator Tube Inspection Program. In Section 2.0, "Proposed Change,"
TSTF-510-A states that references to "tube repair criteria" in Paragraph d is revised to "tube plugging [or repair] criteria." However, in the following sentence in Paragraph d.2, this change was inadvertently omitted, "If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated.
HNP does not have an approved tube repair criteria. Therefore, the sentence is revised to state "tube plugging" criteria.
3.0 Regulatory Evaluation 3.1 No Significant Hazards Consideration Determination Duke Energy requests adoption of an approved change to the standard technical specifications (STS) into the plant specific technical specifications (TS), to revise the Specification 6.8.4.l, "Steam Generator (SG) Program," 6.9.1.7, "Steam Generator Tube Inspection Report," and TS 3.4.5, "Steam Generator Tube Integrity," to address inspection periods and other administrative changes and clarifications.
As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:
- 1.
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plants licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions.
Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
U.S. Nuclear Regulatory Commission Page 3 of 4 HNP-14-050 Enclosure
- 2.
Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated?
Response: No.
The proposed changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.
Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3.
Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary systems pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.
Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Duke Energy concludes that the proposed amendment(s) does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of "no significant hazards consideration" is justified.
4.0 Environmental Evaluation The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or 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 change 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 change.
U.S. Nuclear Regulatory Commission Page 4 of 4 HNP-14-050 Enclosure 5.0 References 5.1 Technical Specifications Task Force Change TSTF-510-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," dated October 11, 2010 5.2 Federal Register Volume 76, Number 208, Page 6676 (76 FR 66763),dated October 27, 2011 5.3 Technical Specifications Task Force Correction to TSTF-510-A, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," dated March 28, 2012
HNP-14-050 Enclosure Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Evaluation of the Proposed Change Application to Revise Technical Specifications to Adopt TSTF-510-A, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process Proposed Technical Specification Changes (7 pages plus cover)
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR CSG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 Steam generator tube integrity shall be maintained.
AND All steam generator tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.
APPLICABILITY:
MODES 1. 2. 3. and 4.
ACTION':
- a.
With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Steam Generator Program:
- 1.
Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection.
and *
- 2.
Plug the affected tube(s) in accordance with the Steam Generator Program prior tO entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.
AND
- b.
With the requirements and associated allowed outage time of ACTION a.. above. not met or SG tube integrity not maintained. be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.5.1 4.4.5.'2 Verify steam generator tube integrity in accordance with the Steam Generator Program.
Verify that each inspected steam generator tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a steam generator tube inspection.
- Separate ACTION entry is allowed for each SG tube.
SHEARON HARRIS - UNIT 1 3/4 4-13 Amendment No. 1 2 4 plugging plugging plugging
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- l. Steam Generator CSG) Program A Steam Generator Program shall be established and implemented to ensure that steam generator CSG) tube integrity is maintained.
In addition. the Steam Generator Program shall include the following provisions:
- 1.
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 a 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.
- 2.
Performance criteria for SG tube integrity. Steam generator tube integrity shall be maintained by meeting the performance criteria for tube structural integrity. accident induced leakage. and operational leakage.
a) Structural integrity performance criterion. All inservice SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup.
operation in the power range. HOT STANDBY. and cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 (3deltaP) 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.
b) 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.
Accident induced leakage is not to exceed 1 gpm total for all three SGs.
c)
The operation leakage performance criterion is specified in LCO 3.4.6.2. "Reactor Coolant System Operational Leakage."
SHEARON HARRIS - UNIT 1 6-19d Amendment No. 124
),
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 3.
Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
- 4.
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 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 4a. 4b. and 4c 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 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.
a)
Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
b)
Inspect 100% of the tubes at sequential periods of 144. 108.
- 72. and thereafter. 60 effective full power months.
The first sequential period shall be considered to begin after the first inservice inspection of the SGs.
In addition.
inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period.
No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.
SHEARON HARRIS - UNIT 1 6-19e Amendment No. 124 plugging plugging assessment installation
<INSERT A>
<INSERT A>
After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in 1), 2), 3), and 4) below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected 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 number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.
- 1) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period;
- 2) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period;
- 3) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and
- 4) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) c) If crack indications are found in any SG tube. then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less).
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.
- 5.
Provisions for monitoring operational primary-to-secondary leakage.
SHEARON HARRIS - UNIT 1 6-19f Amendment No. 124 affected and potentially affected results in more frequent inspections
ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient. 3.1.3.5 - Shutdown Bank Insertion Limits. 3.1.3.6 -
Control Bank Insertion Limits. 3.2.1 -Axial Flux Difference.
3.2.2 - Heat Flux Hot Channel Factor. and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- o.
Mechanical Design Methodologies XN-NF-81-58(P)(A). "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model." approved version as specified in the COLR.
AI~F-81-58(P)(A). "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model." approved version as specified in the COLR.
XI~-NF-82-06(P)(A). "Qualification of Exxon Nuclear Fuel for Extended Burnup." approved version as specified in the COLR.
ANF-88-133(P)(A). "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU." approved version as specified in the COLR.
XI~-NF-85-92(P)(A). "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results."
approved version as specified in the COLR.
EMF-92-ll6(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs." approved version as specified in the COLR.
(Methodologies for Specification 3.2.1 -Axial Flux Difference.
3.2.2 - Heat Flux Hot Channel Factor. and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g.. fuel thermal-mechanical limits. core thermal-hydraulic limits. nuclear limits such as shutdown margin. and transient and accident analysis limits) of the safety analysis are met.
6.9.1.6.4 The CORE OPERATING LIMITS REPORT. including any mid-cycle revisions or supplements. shall be provided. upon issuance for each reload cycle. to the NRC Document Control Desk. with copies to the Regional Administrator and Resident Inspector.
6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following compl~tion of an inspection performed in accordance with Specification 6.8.4.1.
fhe report shall include:
- a.
The scope of inspections performed on each SG.
- b.
Active degradation mechanisms found.
- c.
Nondestructive examination techniques utilized for each degradation mechanism.
SHEARON HARRIS - UNIT 1 6-24c Amendment No. 135 D
ADMINISTRATIVE CONTROLS 6 9 l 7 STEAM GENERATOR TUBE INSPECTION REPORT CCont1nued)
- d.
Locat1on. or1entation (if linear). and measured sizes C1f ava1lable) of service 1nduced indications.
- e.
Number of tubes plugged dur1ng the 1nspect1on outage for each active degradat1on mechanism.
- f.
Total number and percentage of tubes plugged to date. and
- g.
The results of cond1tion monitoring. including the results of tube pulls and 1n-situ test1ng.
SPECIAL REPORTS 6.9.2 Special reports shall be subm1tted to the NRC in accordance with 10CFR50.4 within the time period specified for each report.
6.10 DELETED (PAGE 6-25 DELETED By Amendment No.92)
SHEARON HARRIS - UNIT l 6-24d Amendment No.135 The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator
HNP-14-050 Enclosure Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Evaluation of the Proposed Change Application to Revise Technical Specifications to Adopt TSTF-510-A, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process Revised Technical Specification Pages (6 pages plus cover)
Separate ACTION entry is allowed for each SG tube.
SHEARON HARRIS - UNIT 1 3/4 4-13 Amendment No.
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 Steam generator tube integrity shall be maintained.
AND All steam generator tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION*:
- a.
With one or more SG tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program;
- 1.
Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and
- 2.
Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.
AND
- b.
With the requirements and associated allowed outage time of ACTION a., above, not met or SG tube integrity not maintained, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify steam generator tube integrity in accordance with the Steam Generator Program.
4.4.5.2 Verify that each inspected steam generator tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a steam generator tube inspection.
SHEARON HARRIS - UNIT 1 6-19d Amendment No.
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- l.
Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that steam generator (SG) tube integrity is maintained. In addition, the Steam Generator Program shall include the following:
- 1.
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 a 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.
- 2.
Performance criteria for SG tube integrity. Steam generator tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.
a)
Structural integrity performance criterion. All inservice SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, HOT STANDBY, and cooldown), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 (3deltaP) 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.
b)
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. Accident induced leakage is not to exceed 1 gpm total for all three SGs.
c)
The operation leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."
SHEARON HARRIS - UNIT 1 6-19e Amendment No.
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 3.
Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
- 4.
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 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 plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of 4a, 4b, and 4c 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.
a)
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
b)
After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in 1), 2), 3), and 4) below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected 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 number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.
- 1) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months.
This constitutes the first inspection period;
- 2) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period;
- 3) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and
SHEARON HARRIS - UNIT 1 6-19f Amendment No.
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 4) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.
c)
If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). 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.
- 5.
Provisions for monitoring operational primary-to-secondary leakage.
SHEARON HARRIS - UNIT 1 6-24c Amendment No.
ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -
Nuclear Enthalpy Rise Hot Channel Factor).
- o.
Mechanical Design Methodologies XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," approved version as specified in the COLR.
ANF-81-58(P)(A), "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," approved version as specified in the COLR.
XN-NF-82-06(P)(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup,"
approved version as specified in the COLR.
ANF-88-133(P)(A), "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU," approved version as specified in the COLR.
XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.
EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs,"
approved version as specified in the COLR.
(Methodologies for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.
6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.l. The report shall include:
- a.
The scope of inspections performed on each SG,
- b.
Degradation mechanisms found,
- c.
Nondestructive examination techniques utilized for each degradation mechanism,
(PAGE 6-25 DELETED By Amendment No.92)
SHEARON HARRIS - UNIT 1 6-24d Amendment No.
ADMINISTRATIVE CONTROLS 6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT (Continued)
- d.
Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e.
Number of tubes plugged during the inspection outage for each degradation mechanism,
- f.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and
- g.
The results of condition monitoring, including the results of tube pulls and in-situ testing.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10 CFR 50.4 within the time period specified for each report.
6.10 DELETED
HNP-14-050 Enclosure Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Evaluation of the Proposed Change Application to Revise Technical Specifications to Adopt TSTF-510-A, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process Proposed Technical Specification Bases Changes For Information Only (4 pages plus cover)
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (Continued)
Applicable Safety Analyses The steam generator tube rupture (SGTR) accident is the limiting design 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 1 gallon per minute (gpm). plus the leakage rate associated with a double-ended rupture of a single tube. The accident radiological analysis for a SGTR assumes the ruptured SG secondary fluid is released directly to the atmosphere due to a failure of the PORV in the open position.
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 some analyses developed by the industry. the steam discharge to the atmosphere is based on the total primary-to-secondary leakage from all SGs of 1 gpm. or is assumed to increase to 1 gpm as a result of accident induced conditions. The HNP accident analyses assume the amount of primary-to-secondary SG tube leakage is 1 gpm. For accidents that do not involve fuel damage. the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the limits in LCO 3.4.8. "Reactor Coolant System Specific Activity." 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 10 CFR 50.67 (Reference 2).
SG tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
Limiting Condition for Operation CLCO)
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.
SHEARON HARRIS - UNIT 1 B 3/4 4-2d Amendment No. 124 I plugging plugging plugging
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (continued ACTIONS a.1 and a.2 ACTIONS a.1 and a.2 apply 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 Surveillance Requirement 4.4.5.2. An evaluation of SG tube integrity of the affected tube(s) must be made. SG 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. ACTION b. applies.
An allowed completion time of seven days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.
If the evaluation determines that the affected tube(s) have tube integrity. ACTION a.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.
However. the affected tube(s) must be plugged prior to entering HOT SHUTDOWN following the next refueling outage or SG inspection. This allowed completion time is acceptable since operation until the next inspection is supported by the operational assessment.
ACTION b.
If the requirements and associated completion time of ACTION a.
are not met or if SG tube integrity is not being maintained. the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The allowed completion times are reasonable. based on operating experience. to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SHEARON HARRIS - UNIT 1 B 3/4 4-2g Amendment No. 124 I plugging plugging
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR CSG) TUBE INTEGRITY (continued)
Surveillance Requirements 4.4.5.1 During shutdown periods. the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06. Steam Generator Program Guidelines (Reference 1). and its referenced EPRI Guidelines. establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.
During SG inspections. a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.
The Steam Generator Program determines the scope of the inspection and the method used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope Ci.e.. which tubes or area of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology. nondestructive examination CNDE) technique capabilities. and inspection locations.
The Steam Generator Program defines the frequency of SR 4.4.5.1. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Reference 5). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition. specification 6.8.4.1 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.
SHEARON HARRIS - UNIT 1 B 3/4 4-2h Amendment No. 124 I If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by specification 6.8.4.l until subsequent inspections support extending the inspection interval.
plugging
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY (continued) 4.4.5.2 During an SG inspection. any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in specification 6.8.4.1 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition. the tube repair criteria. in conjunction with other elements of the Steam Generator Program. ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.
The frequency of "Prior to entering HOT SHUTDOWN following a SG inspection" ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary presssure differential.
References
- 1. NEI 97-06. "Steam Generator Program Guidelines"
- 2. 10 CFR 50.67
- 3. ASME Boiler and Pressure Vessel Code.Section III. Subsection NB
- 4. Draft Regulatory Guide 1.121. "Basis for Plugging Degraded Steam Generator Tubes." August 1976
- 5. EPRI. "Pressurized Water Reactor Steam Generator Examination Guidelines" SHEARON HARRIS - UNIT 1 B 3/4 4-2i Amendment No. 124 plugging plugging plugging plugging