ET 22-0006, Operating Corp. - Application to Revise Technical Specifications to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections

Operating Corp. - Application to Revise Technical Specifications to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections

ML22335A570
Person / Time
Site:	Wolf Creek
Issue date:	12/01/2022
From:	Boyce M Wolf Creek
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
ET 22-0006
Download: ML22335A570 (1)

Text

Michael T. Boyce Vice President Engineering December 1, 2022 ET 22-0006 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Docket No. 50-482: Application to Revise Technical Specifications to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections Commissioners and Staff:

Pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting a request for an amendment to the Technical Specifications (TS) for the Wolf Creek Generating Station (WCGS).

WCNOC requests adoption of Technical Specification Task Force (TSTF)-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections, which is an approved change to the Standard Technical Specifications (STS), into the WCGS TS. The TS related to steam generator (SG) tube inspections and reporting are revised based on operating history.

Attachment I provides a description and assessment of the proposed changes. Attachment II provides the existing TS pages marked to show the proposed changes. Attachment III provides revised (clean) TS pages. The TS Bases are not affected by the proposed changes.

WCNOC requests that the amendment be reviewed under the Consolidated Line Item Improvement Process (CLIIP). Approval of the proposed amendment is requested within 12 months of completion of the NRCs acceptance review. Once approved, the amendment shall be implemented within 60 days of completion of 100% inspection of each steam generator.

There are no regulatory commitments made in this submittal.

In accordance with 10 CFR 50.91, Notice for public comment, State consultation, Section (b)(1),

a copy of this application, with Attachments, is being provided to the designated Kansas State official.

ET 22-0006 Page 2 of 3 If you have any questions concerning this matter, please contact me at (620) 364-8831 x8687, or Dustin Hamman at (620) 364-4204.

Sincerely, Michael T. Boyce MTB/

Attachments: I Evaluation of Proposed Change II Proposed Technical Specification Changes (Mark-Up)

III Revised Technical Specification Pages cc:

S. S. Lee (NRC), w/a S. A. Morris, (NRC), w/a K. S. Steves (KDHE), w/a G. E. Werner (NRC), w/a Senior Resident Inspector (NRC), w/a

ET 22-0006 Page 3 of 3 STATE OF KANSAS

) ) ss COUNTY OF COFFEY )

Michael T. Boyce, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

By _ ________________ _ _

Michael T. Boyce Vice President Engineering SUBSCRIBED and swam to before me this 1 s~~ 01 1

Dececr:_: _

NOTARY PU8LIC - State of Kansas Nota~

V CINOY NOVI GE

~

My Appt. Expires ~.!.i".c)._~~:,..!,;a,1 Expiration Date __

8:)~8/--__,2,'-"--E,b __ _

Attachment I to ET 22-0006 Page 1 of 4 EVALUATION OF PROPOSED CHANGE

Subject:

Application to Revise Technical Specifications to Adopt TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation 2.2 Optional Changes and Variations

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration 3.2 Conclusion

4.0 ENVIRONMENTAL CONSIDERATION

Attachment I to ET 22-0006 Page 2 of 4

1.0 DESCRIPTION

Wolf Creek Nuclear Operating Corporation (WCNOC) requests adoption of Technical Specification Task Force (TSTF)-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections, which is an approved change to the Standard Technical Specifications (STS),

into the Wolf Creek Generating Station (WCGS) 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 Published Safety Evaluation WCNOC has reviewed the safety evaluation for TSTF-577-A, Revision 1, 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-A, Revision 1. As described herein, WCNOC has concluded that the justifications presented in TSTF-577-A, Revision 1, and the safety evaluation prepared by the NRC staff are applicable to WCGS and justify this amendment for the incorporation of the changes to the WCGS TS.

The current SG TS requirements are based on TSTF-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection. The SG tubes are made from Thermally Treated Alloy 600 (Alloy 600TT).

The initial inspection period described in the proposed TS 5.5.9, Steam Generator (SG)

Program, paragraph d.2, will be performed during Refueling Outage 26 scheduled for Spring 2024 with a 100% inspection in each steam generator.

2.2 Optional Changes and Variations WCNOC is proposing the following variations from the TS changes described in TSTF-577-A, Revision 1, or the applicable parts of the NRC staffs safety evaluation:

The current WCGS TS 5.5.9d.3, first sentence, the word affect is revised to affected for consistency with the STS and TSTF-577-A, Revision 1.

The WCGS TS utilize different numbering than the Standard Technical Specifications on which TSTF-577-A, Revision 1, was based. Specifically, the Steam Generator Tube Inspection Report is Specification 5.6.10 in the WCGS TS instead of Specification 5.6.7.

The current WCGS TS 5.6.10 first paragraph references the title of TS 5.5.9, Steam Generator (SG) Program, without placing the title in quotes. For consistency with TSTF-577-A, Revision 1, and the STS, the title is placed in quotes.

These changes are administrative and do not affect the applicability of TSTF-577-A, Revision 1, to the WCGS TS.

The WCGS SG Program TS currently contains a provision for an alternate tube plugging criteria.

The description of the alternate tube plugging criteria in the proposed change is equivalent to the description in the current TS.

Attachment I to ET 22-0006 Page 3 of 4

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Wolf Creek Nuclear Operating Corporation (WCNOC) requests adoption of TSTF-577-A, Revision 1, Revised Frequencies for Steam Generator Tube Inspections, which is an approved change to the Standard Technical Specifications (STS), into the Wolf Creek Generating Station (WCGS)

Technical Specifications (TS). The TS related to steam generator (SG) tube inspections and reporting are revised based on operating history.

WCNOC 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 for the current and subsequent inspection intervals. 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 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.

Attachment I to ET 22-0006 Page 4 of 4

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 Updated Safety Analysis Report (USAR) or the operating license.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, WCNOC 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 Commissions 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 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or 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.

Attachment II to ET 22-0006 Page 1 of 8 Proposed Technical Specification Changes (Mark-Up)

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-11 Amendment No. 123, 159, 164, 199 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program An SG Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Steam 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.

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 steam 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), 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.

Leakage is not to exceed 1 gpm per SG.

(continued)

Attachment II to ET 22-0006 Page 2 of 8

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-12 Amendment No. 123, 153, 172, 178, 186, 195, 199, 201 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.13, RCS 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:

1.

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 portion of the tube below 15.21 inches from the top of the tubesheet is excluded 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.

2.

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 Attachment II to ET 22-0006 Page 3 of 8

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-13 Amendment No. 123, 153, 172, 178, 186, 195, 199, 201 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. After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other 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 (continued)

Attachment II to ET 22-0006 Page 4 of 8

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-14 Amendment No. 123, 153, 172, 178, 186, 195, 199, 201 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued) each inspection period as defined in a, b, and c 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.

a)

After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period.

b)

During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c)

During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.

3.

If crack indications are found in any portion of the SG tube excluding any region that is exempt from inspection by alternate repair criterianot excluded above, 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 not exceed 24 effective full power months or one 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 (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.

Attachment II to ET 22-0006 Page 5 of 8

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-15 Amendment No. 123, 153, 172, 178, 186, 195, 199, 201

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

(continued)

Attachment II to ET 22-0006 Page 6 of 8

Reporting Requirements 5.6 Wolf Creek - Unit 1 5.0-28 Amendment No. 123, 142, 158, 164, 178, 179, 186, 195, 199, 201 5.6 Reporting Requirements 5.6.10 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, 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;

b.

Degradation mechanisms found;

c.

Nondestructive examination techniques utilized for each degradation mechanism;

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; ef.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each SGsteam generator;

f.

The results of any SG secondary side inspections; Attachment II to ET 22-0006 Page 7 of 8

Reporting Requirements 5.6 Wolf Creek - Unit 1 5.0-28 Amendment No. 123, 142, 158, 164, 178, 179, 186, 195, 199, 201

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing; gh.

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; hi.

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 2.50 times the maximum operational primary to secondary leak rate, the report should describe how it was determined; and ij.

The results of monitoring for the tube axial displacement (slippage). If slippage is discovered, the implications of discovery and corrective action shall be provided.

Attachment II to ET 22-0006 Page 8 of 8

Attachment III to ET 22-0006 Page 1 of 10 Revised Technical Specification Pages

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-11 Amendment No. 123, 159, 164, 199, 5.5 Programs and Manuals 5.5.9 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.

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 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 per SG.

(continued)

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-12 Amendment No. 123, 153, 172, 178, 186, 195, 199, 201, 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.13, RCS 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:

1.

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.

Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.

2.

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 (continued)

Programs and Manuals 5.5 Wolf Creek - Unit 1 5.0-13 Amendment No. 123, 153, 172, 178, 186, 195, 199, 201, 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

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.

3.

If crack indications are found in any 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.

(continued)

Reporting Requirements 5.6 Wolf Creek - Unit 1 5.0-28 Amendment No. 123, 142, 158,164, 178, 179, 186, 195, 199, 201, 5.6 Reporting Requirements 5.6.10 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, 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 technique utilized;

2.

The location, orientation (if linear), measure 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 analysis methodology, inputs, and results;

e.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;

f.

The results of any SG secondary side inspections;

g.

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; (continued)

Reporting Requirements 5.6 Wolf Creek - Unit 1 5.0-29 Amendment No.

5.6 Reporting Requirements 5.6.10 Steam Generator Tube Inspection Report (continued)

h.

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 2.50 times the maximum operational primary to secondary leak rate, the report should describe how it was determined; and

i.

The results of monitoring for the tube axial displacement (slippage). If slippage is discovered, the implications of discovery and corrective action shall be provided.

High Radiation Area 5.7 Wolf Creek - Unit 1 5.0-30 Amendment No. 123, 142, 158, 164,

179, 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation:

a.

Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b.

Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c.

Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d.

Each individual or group entering such an area shall possess:

1.

A radiation monitoring device that continuously displays radiation dose rates in the area; or

2.

A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or

3.

A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or

4.

A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (continued)

High Radiation Area 5.7 Wolf Creek - Unit 1 5.0-31 Amendment No. 123, 142, 158, 164, 179, 196, 5.7 High Radiation Area 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation: (continued)

(i)

Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii)

Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, or personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

e.

Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation:

a.

Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

1.

All such door and gate keys shall be maintained under the administrative control of the Shift Manager/Control Room Supervisor or health physics supervision, or his or her designee.

2.

Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

b.

Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

(continued)

High Radiation Area 5.7 Wolf Creek - Unit 1 5.0-32 Amendment No. 123, 142, 158, 164,

179, 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation: (continued)

c.

Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d.

Each individual or group entering such an area shall possess:

1.

A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or

2.

A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or

3.

A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i)

Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii)

Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area, or

4.

In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the As Low As is Reasonably Achievable principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

(continued)

High Radiation Area 5.7 Wolf Creek - Unit 1 5.0-33 Amendment No. 123, 142, 158, 164,

179, 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation: (continued)

e.

Except for individuals qualified in radiation protection procedures or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them.

f.

Such individual areas that are within a larger area, such as PWR containment, where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.