L-06-063, Response to Request for Additional Information Regarding Steam Generator Tube Integrity Technical Specification License Amendment Requests

Response to Request for Additional Information Regarding Steam Generator Tube Integrity Technical Specification License Amendment Requests

ML061210055
Person / Time
Site:	Beaver Valley
Issue date:	04/25/2006
From:	Mende R FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-06-063, TAC MC8861, TAC MC8862
Download: ML061210055 (12)

Text

FENOCP FirstEnergy Nuclear Cperating Company RichardG. Mende 724-682-7773 Director, Site Operations Fax: 724-682-1840 April 25, 2006 L-06-063 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Response to Request for Additional Information Regarding Steam Generator Tube Integrity Technical Specification License Amendment Requests (TAC Nos. MC8861 and MC8862)

By letter dated November 7, 2005, the FirstEnergy Nuclear Operating Company (FENOC) submitted License Amendment Request (LAR) Nos. 324 and 196 that would revise steam generator tube integrity technical specifications for Beaver Valley Power Station Unit Nos. 1 and 2. Subsequently, by letter dated March 1, 2006, the NRC requested further information regarding the FENOC submittals. The FENOC responses to this request are provided in Attachment 1. Attachment 1 also addresses an additional question that FENOC has numbered as Item 16, that was received by e-mail on March I 0, 2006. A supplement to the LAR will be provided by June 9, 2006 to incorporate the proposed LAR modifications described in Attachment 1. Attachment 2 to this letter provides a list of regulatory commitments made in this submittal.

If there are any questions or if additional information is required, please contact Mr. Gregory A. Dunn, Manager - FENOC Fleet Licensing, at (330) 315-7243.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 2*', 2006.

Sincerely, Richard G. Mende

%kO

Beaver Valley Power Station, Unit Nos. 1 and 2 Response to RAI on LAR Nos. 324 and 196 L-06-063 Page 2 Attachments

1. FENOC Response to Request for Additional Information

2. Commitment List

References:

1. B-aver Valley Unit Nos. 1 and 2 License Amendment Request Nos. 324 and 196 -

Si:eam Generator Tube Integrity, dated November 7, 2005

2. B saver Valley Power Station, Unit No. 2 - Request for Additional Information -

Regarding The Steam Generator Tube Integrity Technical Specification License Amendment Request (TAC Nos. MC8861 and MC8862), dated March 1, 2006 c: Mr. T. G. Colburn, NRR Senior Project Manager Mr. P. C. Cataldo, NRC Senior Resident Inspector Mr. S. J. Collins, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mor. L. E. Ryan (BRP/DEP)

Attachment 1 to L-06-063 FENOC Response to Request for Additional Information Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2)

Steam Generator (SG) Tube Integrity Technical Specification (TS)

License Amendment Request (LAR)

Docket Nos. 50-334 and 50-412 By letter dated November 7,2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML053140195), FirstEnergy Nuclear Operating Company (FENOC or the licensee) submitted an LAR regarding BVPS-1 and 2 SG tube integrity TSs. The proposed amendment would revise the SG tube integrity TSs to be consistent with the Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4 (ADAMS Accession No. ML051090200).

The NR'C staff has determined that the additional information contained in the enclosure to this letter is needed to complete its review.

1. Proposed Limiting Condition for Operation (LCO) 3.4.5.b for both BVPS-1 and 2 states, "With Action a not being completed within the specified completion time, be iin 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 following

30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />." TSTF-449 states the following for this specific LCO

"If the Required Actions and associated Completion Times of Condition A are not met or if steam generator (SG) tube integrity is not being maintained, the reactor must be brought ito MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />." Please provide justification for removing the key requirement to shutdown the reactor if SG tube integrity is not being maintained or alternatively discuss your plans to modify your TS LCO to include this key requirement and be consistent with TSTF-449. Also, discuss why you elected to remove the modes for this specification but not for others l(i.e., you specify HOT STANDBY rather than MODE 3).

Response

Consistent with TSTF449, the proposed technical specification revisions will be modified to include a requirement to bring the reactor to Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> when steam generator tube integrity is not being maintained.

Mode descriptions were initially used in lieu of Mode numbers to maintain the style of the current TS, much like the timing requirements of the TSTF ("within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />", for example) were adapted to match the current TS style (revised to "within the next

30 hours"), without affecting its technical meaning. In the case of the timing adaptation,

Attachment I to L-06-063 Page 2 it was important to maintain CTS style so that the time period would not be misinterpreted in light of past practice. However, Mode terminology and Mode numbers are clearly defined in the TS and are used interchangeably by plant staff. The use of differing styles was not intended to convey any difference in meaning. However, LCO 3.4.5.b will be modified to use Mode numbers in lieu of the originally proposed termninology.

2. Proposed Surveillance Requirement (SR) 4.4.5.1 for both BVPS-1 and 2 states that the SG Program is used to verify SG tube integrity at a SG tube inspection frequency specified in the SG Program. Given that the SG Program only provides maximum inspection intervals, this statement is not appropriate. In addition, the maximum intervals provided in the SG Program may not be sufficient to ensure SG tube integrity and therefore, it may be necessary to inspect more frequently to ensure that SG tube integrity is being maintained. Please discuss your plans to remove the statement regarding the SG tube inspection frequency.

'Response The proposed changes to technical specification 4.4.5.1 will be modified to remove wording about the frequency for verifying SG tube integrity.

3. On page 3 of your November 7,2005, submittal (ADAMS Accession No.

M\WL0531401950) you indicated that for BVPS-1 your current and proposed TS operational primary-to-secondary leakage limit is 150 gallons-per-day (gpd) per 1SG measured at room temperature conditions. This leak rate (i.e., 150 gpd) is also what

!is assumed in your design-basis accident (DBA) analysis.

The NRC staff and the industry (through TSTF-449, Revision 4) have used the term accident-induced leakage to include any primary-to-secondary leakage existing prior to the accident plus the primary-to-secondary leakage induced during the accident. This was done, in part, because with today's technology it is not possible to distinguish whether the leakage during a DBA is coming from flaws that were leaking during normal operation or whether the leakage is coming from flaws that were not leaking during normal operation. Based on your proposed TS Bases section 3/4.4.5, "Steam Generator (SG) Tube Integrity," it appears that you have adopted this definition of accident-induced leakage.

The NRC staff recognizes that plants have assumed that the leak rate during a D BA is the same as the leak rate during normal operation. However, it is important and required to ensure that neither of these limits are exceeded. As a result, it may be necessary to ensure that the operational leak rate is kept well below the operational leak rate limit since the leak rate experienced during a DBA may be higher than that observed during normal operation. This increase in leak rate can be a result of either, (1) the higher differential pressure associated with a DBA causing the leak rate from flaws leaking during normal operation to leak higher rates or (2) the higher loadings associated with a DBA causing a flaw that was not leaking during normal operation to leak during the accident.

Attachment 1 to L-06-063 Page 3 Although BVPS-1 plans to replace the existing SGs with new SGs having a number of improved design and material changes in the spring of 2006, it is possible in the future that you will observe operating leakage and that you may also be postulating leakage during a DBA. From your submittal, it is not clear that you won't exceed the accident-induced leakage limit. For example, if you were projecting to observe an accident-induced primary-to-secondary leak rate of 40 gpd and you had a 60-gpd operational primary-to-secondary leak rate, the leak rate during DBAs could be greater than 150 gpd. This is because the operational primary-to-secondary leak rate may double to 120 gpd due to the increase in the differential pressure associated with the accident. When this leakage is combined with your projected accident-induced leak rate of 40 gpd, the total leak rate during the DBA would be 160 gpd which is greater than that assumed during your DBA analysis. The NRC staff notes that it is most likely not feasible (with today's technology) to ascertain whether the operating leak rate is a result of flaws also projected to leak during a DBA.

'Given the above, discuss whether your procedures recognize this potential leakage lissue or discuss your plans to modify your procedures to ensure that you will not exceed the accident-induced leak rate limit as a result of the higher leak rates that may be observed during a DBA (as a result of inducing "new" leakage or as a result of the higher driving force for leakage). Alternatively, discuss your plans (and the technical basis) for modifying your normal operating and accident-induced leakage Ilimit to address these effects.

Response

Abnormal Operating Procedure 1OM-53C.4.1.6.4, "Steam Generator Tube Leakage" addresses the possibility that primary-to-secondary leakage may increase as a result of accident conditions by initiating plant shutdown in advance of reaching the accident

analysis assumed primary-to-secondary leak rate of 150 gpd. The procedure provides actions to be taken in the event of steam generator tube leakage greater than 5 gpd. In addition to steps that directly address plant conditions, the procedure provides for ongoing monitoring of the leakage and an escalating series of responses depending on the magnitude of the leakage. The procedure requires plant shutdown to Mode 3 within

'24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when leakage of 75 gpd or greater occurs for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If a subsequent increase in leak rate of 30 gpd per hour occurs, the procedure calls for reduction to less than 50 percent power within one hour and shutdown to Mode 3 within the following 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4. Proposed TS Section 6.9.7, "Steam Generator Tube Inspection Report," indicated that you use the Electric I'ower Research Institute (EPRI) Guidelines definition of active degradation. The NRC staff has found that the industry's definition of active degradation is misleading since tubes could have degradation that is progressing (or present on the tubes) but the degradation could be classified as "not active" (refer to ADAMS Accession Nos. ML010320218 and ML012200349). As a result, please

Attachment 1 to L-06-063 Page 4 discuss your plans to modify your TSs to remove the reference to the EPRI Guidelines.

Response

'Consistent with TSTF-449, the proposed technical specification revisions will be modified to remove the reference to the EPRI guidelines definition of active degradation.

5. Given that the new TSs provided in TSTF-449 do not allow operation when the accident-induced leakage criteria is exceeded, please discuss your plans to omit TS Section 6.9.7.1.

'Response Proposed BVPS-2 Technical Specification 6.9.7.1 will be deleted because the proposed rs would not allow operation when the accident-induced leakage performance criteria of the SG program are exceeded.

6. The accident-induced leakage performance criteria in TSTF-449 consists of 2 limits:

(1) a limit established in your current design and licensing basis and (2) a limit of I gpm which is based on severe accident considerations. In your proposal, you replaced the 1-gpm limit with the limit listed in LCO 3.4.6.2.c of 150 gpd for both

BVPS-1 and 2. Although the 150-gpd limit is acceptable (since it is less than 1 gpm),

lit is not clear why it is necessary to reference LCO 3.4.6.2.c. Discuss your plans to remove reference to LCO 3.4.6.2.c and to change the second part of the accident-iinduced leakage limit to 1 gpm, consistent with TSTF-449.

Response

The proposed TS will be modified to remove the reference to LCO 3.4.6.2.c and to replace the 150 gpd value with 1 gpm. Because the wording provided in TSTF-449 does not clearly convey that two separate limits are included in the accident induced leakage performance criteria, editorial clarifications will also be included in the modifications.

Proposed wording (except for differences related to RAI Item 7) would be similar to:

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 also not to exceed 1 gpm per SG.

7. ]For BVPS-2, you proposed in TS Section 6.19.c that leakage would not exceed

150 gpd per SG except for specific types of degradation at specific locations.

However, the actual alternate tube repair criteria that may be applied is described in TS Section 6.19.c.1. Please discuss your plans to modify your proposed TSs to mnore precisely reflect the location of the applicable alternate tube repair criteria.

Since proposed TS Section 6.19.c.1 does not specify what sources of accident-induced leakage shall be limited to 150 gpd, discuss your plans to clarify the

Attachment I to L-06-063 Page 5 accident-induced leakage performance criterion. For example, one way to clarif y the performance criterion might be to state, "Leakage from all sources, excluding the leakage attributed to the degradation described in TS Section 6.19.c.1, is not to exceed 150 gpd per SG."

Response

Proposed BVPS-2 Technical Specification 6.19.b.2 will be modified to remove ambiguity regarding which sources of accident-induced leakage would be subject to the severe accident limit of 1 gpm (note that 150 gpd would be changed to 1 gpm based on RAI Item 6). The modifications would also provide a more precise reference regarding thc type of degradation to which the limit would not be applied.

8. One of the purposes of TSTF-449 is to allow licensees to update their TSs to accurately reflect their SG tube integrity program. TS Section 6.19.c indicated that flaws with a percent through-wall depth of 32 percent for ABB Combustion Engineering Tungsten Inert Gas (TIG) Welded Sleeves and 25 percent for Westinghouse Laser Welded Sleeves are to be plugged. It is the NRC staff's

• understandingthat you are plugging on detection, any flaws in sleeved tubes (refer to page 5 of your submittal). Plugging on detection is normally performed when flaws cannot be reliably sized or when the threshold of detection is near the repair limit. Please discuss your plans to modify your proposed TSs to reflect current industry practice of plugging on detection, flaws in sleeved tubes.

In addition to the above, the original qualification of these sleeves assumed that there was no degradation in the parent tube at the sleeve joint (i.e., a flaw at the plugging limit of 40 percent through-wall was not assumed). As a result of this, i t is the NRC staff's understanding that when degradation is detected in a parent tube at

a joint between a tube and sleeve, that tube would be plugged. Please discuss your plans for modifying your proposed TSs to clarify that plugging on detection is required when degradation of the parent tube is detected at a sleeve joint.

'Response

Refer to the response provided for RAI Item 15.

9. The level of detail provided in proposed TS Section 6.19.d.4, regarding sleeve inspections is no longer needed. This information is superceded by the requirement ito inspect the tubes with eddy current techniques and equipment at intervals capable of detecting all possible flaw types and at intervals necessary to ensure SGJ itube integrity. Please discuss your plans to remove TS Section 6.19.d.4.

'Response Proposed TS 6.19.d.4 will be removed because it is superseded by broader requirements governing sampling, scope and frequency of inspections that are described elsewhere in the proposed SG program TS.

Attachment I to L-06-063 Page 6

10. In order for your TSs to accurately reflect your SG tube integrity program, TS Section 6.19.d.5 should indicate that the tubes will be inspected full-length with a bobbin probe in accordance with the tube-to-tube support plate repair criteria amendment. Please discuss your plans to further clarify the inspection requirements in TS Section 6.19.d.5. The bobbin probe is specified in your current TSs.

Response

References to the bobbin probe were not included in the proposed TS Section 6.19.d.5 because they were believed to be a level of detail that was no longer needed given the new requirement to inspect the tubes with eddy current techniques capable of detecting all possible flaw types necessary to ensure SG tube integrity. However, the specification will be modified to include specific requirements to use a bobbin coil, consistent with the current TS. In addition, the "100-percent" phrase will be replaced with "full-length" so that the portion of tube requiring inspection is not confused with the quantity of tubes requiring inspection.

11. A safety factor of 1.4 against burst applied to the DBA primary-to- secondary pressure differentials was indicated in TS Section 6.19.b.1. Generic Letter (GL) !5-0D5, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," indicated that there is a possibility that a tube may have a burst pressure less than 1.4 times the steam line break pressure differential (given the uncertainties associated with the various correlations), therefore, the GL 95-05 alternate repair criteria (ARC) imposed a limit on the probability of burst (POB) of 1X10-2 . As a result, it is not clear from your submittal that the structural integrity performance criteria is complete, since it does not fully address all the performance criteria for implementation of the voltage-based ARC. Please discuss your plans to modify the performance criteria to Fully address the voltage-based ARC. For example, discuss your plans for modifying the structural integrity performance criteria to indicate that for predominately axially-oriented outside diameter stress-corrosion cracking (ODSCC) at the tube support plate elevations, the POB of one or more indications, given a steam line break, shall be less than 1X1O 2.

Response

Proposed TS Section 6.19.b.1 will be modified to include the probability of burst for predominantly axially oriented ODSCC at tube support plate elevations as part of the SG program structural integrity performance criteria. Given that the proposed modification would not allow operation when the POB criterion is exceeded, reporting requirement

'rs 6.9.7.5 would no longer be needed and will be removed from the proposed TS.

12. For BVPS-2, you indicate on page 6 of your submittal that the main steam line break dose analysis assumes 2.8 gpm leakage (2.5 gpm from the faulted SG and 0.3 gpm from the non-faulted SG). You further indicate that other accidents (that assume primary-to-secondary leakage exists) assume there is 150 gpd leakage.

to L-06-063 Page 7 Please clarify whether this is 150 gpd per SG or 150 gpd total from all three SGs. If the latter, please discuss what controls are in place to make sure the accident-induced leakage criteria is not exceeded as a result of having operating leakage from all three SGs at levels below their limit (e.g., all 3 SGs leaking at 60 gpd).

Response

Accidents other than main steam line break assume primary-to-secondary leakage of 150 gpd per SG (i.e., 450 gpd for all), consistent with the current TS.

13. In proposed TS Section 6.9.7.a for BVPS-2, you indicate that you will also report the number and extent of sleeves examined. Since this should already be included in the "scope of inspections performed on each SG," it is not clear why this was added.

Please discuss why this extra phrase is needed (i.e., not encompassed in the first part of the requirement) or discuss your plans for removing it.

Response

'The proposed reporting requirement in TS Section 6.9.7.a will be modified to remove the redundant phrase regarding examination of sleeves.

14. The NRC staff has made several other observations regarding TS Section 6.19 for BVPS-2 and they are listed below.

a. TS Section 6.19.c.1 states that the plugging (repair) limit at tube support plate intersections is based on maintaining SG tube serviceability. Please discuss your plans to remove the phrase "maintaining SG tube serviceability" since serviceability is not defined in your proposed TSs.

Response

Proposed TS Sections 6.19.c.1 will be modified to remove the undefined terminology.

lb. Please discuss why the phrase "indications of potential degradation" was removed from TS Sections 6.19.c.1.c and 6.19.c.1.d or alternatively discus,;

your plans to modify your proposed TSs to be consistent with TSTF-449.

Response

The proposed TS Sections 6.19.c.1.c and 6.19.c.1.d will be modified to replace the term "degradation" with the phrase "indications of potential degradation" that appears in the current TS.

c. There are two TS Sections 6.19.c.1.d in your proposed TSs. The second TS Section 6.19.c.1.d should be 6.19.c.1.e. In addition, once the numbering is corrected, TS Section 6.19.c.1.e should state that during an unscheduled mid-cycle inspection, the mid-cycle repair limits apply instead of the limits

Attachment 1 to L-06-063 Page 8 specified in 6.19.c.1.a, 6.19.c.1.b, 6.19.c.1.c, and 6.19.c.1.d. Also, a statement such as, "implementation of these mid-cycle repair limits" should follow the same approach as in TS Sections 6.19.c.1.a through 6.19.c.1.d consistent with your current TSs.

Response

The second proposed TS Section 6.19.c.1.d will be modified to correct the numbering error, to add the missing reference to 6.19.c.1.d, and to add a statement regarding the approach to implementation of the mid-cycle repair limits that is consistent with the current TS.

d. In TS Section 6.19.d, it is stated that previous defects or imperfections in an area repaired by sleeving are not considered an area requiring re-inspection.

Given that the terms "defect" and "imperfection" are not defined in TSTF-449, please discuss your plans to modify this proposed TS using terminology such as, "the portion of the original tube wall between a sleeve's joint does not need to be inspected."

Response

Proposed TS Section 6.19.d will be modified to remove the undefined terms and to substitute wording that directly describes the portion of tube that does not need to be inspected when a sleeve has been installed.

e. TS Section 6.19.c.1.c should be referenced in TS Section 6.19.c1.b.

Response

Proposed TS Section 6.19.c.1.b will be modified to include a reference to 6.19.c.1.c.

15. In your proposed TS (and TSTF-449) a SG tube is defined as the entire length of the tube, including the tube wall and any repairs made to it, between the tube-to-itubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.

Given this definition, the proposed repair criteria in TS 6.19.c could be misinterpreted. Please discuss your plans to modify your proposed TS 6.19.c to more clearly define the repair criteria for the sleeved portion of a tube. For example, the TS may be modified using the following:

a. The non-sleeved region of a tube found by inservice inspection to contain flaws with a depth equal to or exceeding 40 percent of the nominal tube wall thickness shall be plugged or repaired except when alternate tube repair criteria permitted by TSs are satisfied.

Tubes shall be plugged if the sleeved region of a tube is found by in-service inspection to contain flaws in the (a) sleeve or (b) pressure boundary portion to L-06-063 Page 9 of the original tube wall in the sleeve-tube assembly (i.e., the sleeve-to-tube joint).

b. The following alternate repair criteria may be applied as an alternative to the 40-percent depth-based criteria.

Response

Proposed TS Section 6.19.c will be revised to address repair criteria at all points along the tube length that provide a pressure boundary, including the original tube in non-sleeved areas, the repair sleeve, and both of these where a tube to sleeve joint provides the pressure boundary.

16. (Question received via e-mail) One of the purposes of TSTF-449 is to allow licensees to update their TSs to accurately reflect their SG tube integrity program. For implementation of GL 95-05, licensees have submitted "90-day reports" providing information concerning tube pulls and condition monitoring/operational assessment results. Consistent with the philosophy of TSTF-449, please discuss your plans to modify TS Section 6.9.7, Steam Generator Tube Inspection Report, to include a requirement to provide the information described in Section 6b of Attachment 1 of GL 95-05 to the NRC.

Response

Proposed TS Section 6.9.7 for BVPS-2 will be modified to include a requirement to provide the information described in Section 6.b of Attachment 1 of GL 95-05. This reporting requirement would not be applicable to BVPS-1 because voltage-based repair criteria have not been approved for that unit.

ATTACHMENT 2 to L-06-063 Commitment List The following list identifies those actions committed to by FirstEnergy Nuclear Operating Company (FENOC) for Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2 in this document. Any other actions discussed in the submittal represent intended or planned action:, by FENOC. They are described only as information and are not regulatory commitments. Please notify Mr. Gregory A. Dunn, Manager, Fleet Licensing at 330-315-7243 of any questions regarding this document or associated regulatory commitments.

Commitment Due Date

1. A supplement to BVPS-1 LAR 324 and BVPS-2 June 9, 2006 LAR 196 will be provided to incorporate the proposed LAR modifications described in Attachment 1 to Letter L-06-063.