RNP-RA/06-0105, Response to NRC RAI Pertaining to Proposed Technical Specifications Changes Regarding Steam Generator Tube Integrity
| ML063320520 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 11/20/2006 |
| From: | Lucas J Progress Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RNP-RA/06-0105 | |
| Download: ML063320520 (14) | |
Text
&j; Progress Energy Serial: RNP-RA/06-0105 NOV 2 0 2006 United States Nuclear Regulatory Commission ATFN: Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION PERTAINING TO PROPOSED TECHNICAL SPECIFICATIONS CHANGES REGARDING STEAM GENERATOR TUBE INTEGRITY Ladies and Gentlemen:
In a letter dated May 30, 2006, Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. (PEC), requested NRC review and approval of changes to modify the Technical Specifications (TS) requirements related to steam generator tube integrity for H. B.
Robinson Steam Electric Plant (HBRSEP), Unit No. 2. An NRC request for additional information (RAI) pertaining to this amendment request was received by electronic mail transmission dated September 19, 2006. Attachment II to this letter provides the response to the RAI.
Attachment I provides an Affirmation in accordance with the provisions of 10 CFR 50.30(b).
Attachment Ill provides revised edited TS and TS Bases pages. Attachment IV provides the revised TS page.
In accordance with 10 CFR 50.91, a copy of this application is being provided to the State of South Carolina.
If you have any questions concerning this matter, please contact Mr. C. T. Baucom at (843) 857-1253.
Sincerely, Manager - Support Services - Nuclear Progress Energy Carolinas. Inc.
Robinson Nuclear Plant 3581 West Entrance Road Hartsville, SC 29550A 0 1
United States Nuclear Regulatory Commission Serial: RNP-RA/06-0105 Page 2 of 2 JFL/cac Attachments:
I. Affirmation II. Response to NRC Request for Additional Information Pertaining to Proposed Technical Specifications Changes Regarding Steam Generator Tube Integrity III. Revised Edited Technical Specifications and Bases Pages IV. Revised Technical Specifications Page c:
Mr. T. P. O'Kelley, Director, Bureau of Radiological Health (SC)
Mr. H. J. Porter, Director, Division of Radioactive Waste Management (SC)
Dr. W. D. Travers, NRC, Region II Mr. C. P. Patel, NRC, NRR NRC Resident Inspector, HBRSEP Attorney General (SC)
United States Nuclear Regulatory Commission Attachment I to Serial: RNP-RA/06-0105 Page 1 of I AFFIRMATION The information contained in letter RNP-RA/06-0105 is true and correct to the best of my information, knowledge and belief; and the sources of my information are officers, employees, contractors, and agents of Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc. I declare under penalty of perjury that the fo is true and correct.
Executed On:
4 f
___a_,_________________________
William G. Noll Director - Site Operations HBRSEP, Unit No. 2
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/06-0105 Page 1 of 4 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION PERTAINING TO PROPOSED TECHNICAL SPECIFICATIONS CHANGES REGARDING STEAM GENERATOR TUBE INTEGRITY The following responses are provided for the NRC request for additional information (RAI) that was provided in an electronic mail transmission dated September 19, 2006:
NRC Request la:
- 1. In your application, you proposed to reduce the normal operating primary-to-secondary leakage limit from 150 gallons per day (gpd) to 75 gpd. It appears that the primary reason for this reduction was to ensure there was margin between the normal operating leakage limit and the accident induced leakage limit (since a margin between these limits was introduced in TSTF-449). With respect to this issue, please address the following:
- a. In TSTF-449 the normal operating leakage limits were reduced from 1 gallon per minute (gpm) total leakage through all steam generators and 500 gpd leakage through any one steam generator to 150 gpd leakage through any one steam generator. Assuming the accident analysis had the same assumptions, the staff considered the 150 gpd leakage limit significantly less than the conditions assumed in the safety analyses. Given that your accident analysis for a steam generator tube rupture assumes 115 gpd (0.08 gpm) leakage through the faulted steam generator, it does not appear that your proposed normal operating leakage limit of 75 gpd is significantly less than your accident analysis assumptions. As a result, please discuss your plans to modify your Bases (page B 3.4-77) to remove the "significantly" from the following phrase: "...75 gallons per day is significantly less than the conditions assumed in the safety analyses." The staff notes that the proposed 75 gpd normal operating leakage limit still satisfies its original purpose of limiting the frequency of steam generator tube ruptures.
Response
The word "significantly" is not needed and can be removed. A revised edited version of this page is provided in Attachment III.
NRC Request 1b:
- b. In the Limiting Condition for Operation section on page B 3.4-78, you indicate, in part, that the 75 gpd normal operating leakage limit maintains the primary-to-secondary leakage within the applicable accident analysis assumptions. That is, by maintaining your operating leakage less than 75 gpd you will ensure that your accident induced leakage limits will not be exceeded. This wording goes beyond TSTF-449 and is not consistent with analysis of operating experience data.
As a result, please discuss your plans to remove this statement and make your submittal consistent with TSTF-449.
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/06-0105 Page 2 of 4
Response
As stated in the May 30, 2006, submittal of the proposed changes, the typical value for assumed accident leakage rate for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, is 0.3 gpm through the three steam generators (SGs) and 150 gpd through any single SG, as opposed to the 1.0 gpm and 500 gpd leakage rates listed in the generic "No Significant Hazards Consideration."
Therefore, the proposed Technical Specifications (TS) changes include a reduction from 150 gpd to 75 gpd to accommodate this difference. The generic TS change does not provide guidance for this circumstance.
As noted in response to NRC Request Ic, HBRSEP, Unit No. 2, utilizes procedural guidance that accounts for accident induced leakage, which is an additional aspect of providing assurance that accident induced leakage is not expected to exceed analyzed leakage limits. Therefore, the phrase, "... and maintaining primary to secondary leakage within the applicable accident analysis assumptions..." is not necessary and will not be included in the TS Bases. A revised edited version of page B 3.4-78 is provided in Attachment III.
NRC Request 1c:
- c. The NRC staff recognizes that plants have assumed that the leak rate during a design basis accident is the same as the leak rate during normal operation. However, it is important (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 design basis accident 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 design basis accident causing the leak rate from flaws leaking during normal operation to leak at higher rates or (2) the higher loadings associated with a design basis accident causing a flaw that was not leaking during normal operation to leak during the accident.
Given the above, discuss whether you have procedures (or will develop procedures) that will ensure neither of the limits are exceeded. The staff notes that each leakage event may need to be assessed independently since the increase in leakage as a result of accident induced conditions may depend on the flaw mechanisms that could potentially be causing the leak.
Response
As stated in response to NRC Request lb, the typical value for assumed accident leakage rate for HBRSEP, Unit No. 2, is 0.3 gpm through the three SGs and 150 gpd through any single SG, as opposed to the 1.0 gpm and 500 gpd leakage rates listed in the generic "No Significant Hazards Consideration." Therefore, the proposed TS changes include a reduction from 150 gpd to 75 gpd to accommodate this difference. The proposed margin is intended to accommodate accident induced leakage.
Additionally, the procedure EGR-NGGC-0208, "Steam Generator Integrity Program," includes provisions to conduct operational assessments. That procedure states, "The operational assessment is a 'forward looking' prediction of the steam generator tube conditions that is used
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/06-0105 Page 3 of 4 to ensure that the structural integrity and accident leakage performance criteria will not be exceeded during the next operating cycle."
Therefore, the steam generator integrity program procedure has guidance for ensuring that the accident induced leak rate limits are not exceeded.
NRC Request 2:
- 2. Regarding proposed TS 5.5.9.c, you indicate that the repair criteria is: "47% of the nominal tube wall thickness if the next inspection interval of that tube is 12 months, and a 2% reduction in the plugging limit for each 12 month period until the next inspection of the inspected SG."
This wording is consistent with your current technical specifications. Since your proposed technical specifications no longer define the term "plugging limit," please discuss your plans to modify this wording to indicate that the "repair criteria" is reduced based on the inspection interval. In addition, please discuss your plans to clarify that the repair criteria is reduced by 2%
for each 12 month period until the next inspection of "the tube" rather than the "next inspection of the inspected SG" (since there no longer will be a requirement to re-inspect previously degraded tubes in the first random sample of that steam generator). Since the proposed criteria is based on calendar months and the inspection frequencies are based on effective full power months, discuss the need to revise the repair criteria such that it references effective full power months. Alternatively, discuss your plans to make your proposal consistent with other similarly designed and operated units by incorporating a 40% depth based tube repair criteria.
Response
The proposed TS 5.5.9.c states that tubes that contain flaws with depth equal to or exceeding the specified criteria shall be plugged. Therefore, the specified "repair criteria" is a de facto plugging limit. The use of "repair criteria" in lieu of "plugging limit" is considered equivalent.
Therefore, this terminology will be revised.
The requirement to reduce the repair criteria by 2% for each 12 month period (for tubes found with flaw indications) until the next inspection of that tube is consistent with the current TS requirement. Implementation of this requirement provides the method for determining the applicable repair criteria depth. The requirement is specific to the inspection of the tube that contains the flaw. Therefore, it is appropriate to base the reduction on the schedule for the next inspection of the affected tube. This can be clarified by changing the wording to clearly state that the reduction is based on the next inspection of the tube. Therefore, this will be revised. A revised edited version of this page is provided in Attachment III and a revised page is provided in Attachment IV.
As noted in the NRC RAI, the proposed TS requirement for adjustment of the repair criteria is based on calendar months, which is consistent with the current TS requirement. The use of effective full power months would be considered a less restrictive change to the TS and is not being proposed at this time.
United States Nuclear Regulatory Commission Attachment II to Serial: RNP-RA/06-0105 Page 4 of 4 NRC Request 3:
- 3. In the second full paragraph on page B 3.4-113, there appears to be two typographical/
administrative errors: (1) "gpm" should be "gpd" and (2) "except for specific types of degradation at specific locations where the NRC has approved greater accident induced leakage" should be removed since there are no approved alternate repair criteria for Robinson. Please discuss your plans to correct these two typographical/administrative errors.
Response
These errors will be corrected. A revised edited version of this page is provided in Attachment III.
NRC Request 4:
- 4. On page B 3.4-114, you proposed to add the following sentence: "Condition A does not apply to the occurrence of primary to secondary LEAKAGE, which is monitored and maintained in accordance with LCO [Limiting Condition for Operation] 3.4.13." Please discuss why this statement was added.
Response
This sentence was added based on questions that arose during the HBRSEP, Unit No. 2, Plant Nuclear Safety Committee (PNSC) review of the proposed TS changes. Specifically, it was noted by the PNSC that entry into Condition A of LCO 3.4.18 could be required if leakage were to occur in an SG during operation. Based on a review of the technical justification for TSTF-449 it was concluded that Condition A of LCO 3.4.18 is not intended to apply to the discovery of primary to secondary leakage. This additional clarification does not prevent the appropriate use of this condition for the discovery of inspection data that may have met or exceeded the repair criteria or necessitated other evaluations. Therefore, the identified sentence is being added to the basis of Condition A for LCO 3.4.18 to appropriately clarify this condition.
United States Nuclear Regulatory Commission Attachment III to Serial: RNP-RA/06-0105 5 pages (including this page)
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REVISED EDITED TECHNICAL SPECIFICATIONS AND BASES PAGES
Insert:
Revised TS Section 5.5.9 (continued) 41
- 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 $6. Leakage is not to exceed 75 gallons per day per SG.
- 3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RGS Operational LEAKAGE."
- c.
Provisions for SG tube repair criteria-Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding the following criteria shall be plugged:
47% of the nominal tube wall thickness if the next inspection repair interval of that tube is 12 months, and a 2. reduction in the criteria
-*l
" I for each 12 month period until the next inspection of the ingpeeibed S6.4-d-Provisions for SG tube inspections. Periodic S6 tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next S6 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.
1-Inspect 100% of the tubes in each SQ during the first refueling outage following SG replacement.
- 2.
Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the S~s. In
RCS operational Leakage B 3.4.13 BASES (continued)
APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses SAFETY ANALYSES do not address operational LEAKAGE-However, other operational LEAKAGE is related to the safety analyses for LOCA: the amount of leakage can affect the probability of such an event.
The safety analysis for an event resulting that primary to in steam discharge to the atmosphere assumes a 9.; 9p, secondary LEAKAGE 4-P....................
as,,^@ 4........ *t ep.
from all steam generators (SGs) is Primary to secondary LEAKAGE is a factor in the dose 0.3 gpm or increases releases outside containment resulting from a steam line to 0.3 gpm as a break (SLB) accident.
To a lesser extent, other accidents result of accident or transients involve secondary steam release to the induced conditions.
atmosphere, such as a steam generator tube rupture (SGTR).
ided condequirentis The leakage contaminates the secondary fluid.
The [CO requirement to limit primary to For the SGTR, the activity released due to the 0.3 gpm secondary LEAKAGE primary to secondary LEAKAGE is relatively insignificant through any one SG to compared to the activity released via the ruptured tube.
less than or equal to The safety analysis for the SGTR accident assumes 0.3 gpm 75 gallons per day is total primary to secondary LEAKAGE in all generators as an less initial condition.
After mixing in the secondary side, the th fanthe conditions activity is then released via the SG PORVs or safeties.
than the conditions This release pathway continues until the SGs are isolated, assumed in the safety which is relatively soon for the affected SG compared to the analyses.
intact SGs.
The dose consequences resulting from the SGTR accident are within the limits defined in 10 CFR 50.67.
The RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement-LCO RCS operational LEAKAGE shall be limited to:
- a.
Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.
LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE.
Violation of this LCD could result in continued degradation of the RCPB.
LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.
(continued)
HBRSEP Unit No. 2 B 3-4-77 Revision No. -226-
RCS operational Leakage B 3.4.13 BA'1ES (continued)
LCO (continued)
- d. Primary to Secondary LEAKAGE Through Any One SG The limit of 75 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 3).
The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage.
The operational LEAKAGE criterion of 75 gallons per day in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures mid C*,,J ev
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i e
ý M li'i
- b.
Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is alloved as a reasonable minimum detectable amount that the containment atmosphere radiation monitoring systems. condensate measuring system. dewpoint monitoring equipment. and containment sump level monitoring equipment can detect within a reasonable time period.
Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.
- c.
Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of identified LEAKAGE and is well within the capability of the RCS Makeup System.
Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources. but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE).
Violation of this LCD could result in continued degradation of a component or system.
- d.
Primar tt_ econdar EAKAGE thr h All Stea Total rimary to econdary L GE amount' g to 0.3 m throug all SGs pro ces accepta e off'site dos s in the R acciden analysis.
Violation of t s LCO cou exceed th offsite dos limits for is ccident.
rimary to s condary L E must be included - the total lowable lim' for identi ed Prima to Seconda LEAKAGE th uqh Any One/
Th 150 gallons er day lim!
on one SG pr uces a ceptable dos consequence. in the SGTR cident nalysis.
V' lation of th's LCO could e eed the offsite dos limits for t is accident.
imary to secondary KAGE must 1 included in e total allowabl limit for id tified LEA (continued)
HBRSEP Unit No. 2 B 3.4-78 Revision No-Insert New TS Bases Section 3.4.18 SG Tube Integrity B 3.4.18 BASES (Continued)
LCO (continued) significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established" For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads.
For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.
Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III. Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.
This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref.
- 4) and Draft Regulatory Guide 1.121 (Ref. 5).
The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident.
analysis assumptions. The accident analysis ass s
gpd accident induced leakage does not exceed 150_1
. er SG, l*[
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The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.
The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation.
The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE,"
and limits primary to secondary LEAKAGE through any one SG to 75 gallons per day.
This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.
(continued)
HBRSEP Unit No. 2 B 3-4-113 Revi sion No-
United States Nuclear Regulatory Commission Attachment IV to Serial: RNP-RA/06-0105 2 pages (including this page)
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REVISED TECHNICAL SPECIFICATIONS PAGE
Programs and Manuals 5.5 5-5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)
- 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 75 gallons per day per SG.
- 3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
- c.
Provisions for SG tube repair criteria-Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding the following criteria shall be plugged:
47% of the nominal tube wall thickness if the next inspection interval of that tube is 12 months, and a 2% reduction in the repair criteria for each 12 month period until the next inspection of the tube.
- d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and 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 replacement.
- 2.
Inspect 100% of the tubes at sequential periods of 120.
- 90. 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 (continued)
HBRSEP Unit No. 2 5-0-13 Amendment No-