L-HU-06-037, Supplement to Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity
| ML063620460 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/27/2006 |
| From: | Salamon G Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-HU-06-037, TAC MD0209, TAC MD0210 | |
| Download: ML063620460 (87) | |
Text
{{#Wiki_filter:CommiCted to Nuclear Excellence Nuclear Management Company, LLC L-HU-06-037 10 CFR 50.90 December 27,2006 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units Iand 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 Supplement to Application For Technical Specification lmprovement Regarding Steam Generator Tube lnteqrity (TAC Nos. MD0209 and MD0210) References 1) License Amendment Request (LAR) titled, "Application For Technical Specification lmprovement Regarding Steam Generator Tube Integrity", dated February 16,2006, Accession Number ML060480440.
- 2) Supplement to Application For Technical Specification lmprovement Regarding Steam Generator Tube Integrity, dated July 21, 2006. Accession Number ML062370052.
By letter dated February 16,2006 (Reference I ) , Nuclear Management Company (NMC) submitted an LAR to adopt Technical Specification (TS) improvements regarding steam generator tube integrity provided in Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-449, "Steam Generator Tube Integrity", Revision 4. By letter dated July 21, 2006 (Reference 2), NMC submitted proposed TS and Bases changes which replaced in their entirety the changes proposed in Reference I.This letter supplements the referenced LAR to address the Nuclear Regulatory Commission (NRC) Staff requests for additional information (RAls) sent by email on August 29,2006, October 25,2006 and November 9,2006 regarding Enclosures 1 and 2 of Reference 2. NMC is submitting this supplement in accordance with the provisions of 10 CFR 50.90. provides the NRC RAls and NMC responses. Enclosure 2, which includes the TS and Bases pages marked up in response to the RAls, replaces Enclosure 2 of Reference 2 in its entirety. Additions to the current TS and Bases are shown with double-underline and deletions are shown with strikethrough. The proposed changes associated with this supplement appear in Enclosure 2 on pages 5.0-13, 5.0-14, 5.0-20, 5.0-21, 5.0-22, 5.0-26, 5.0-27, 5.0-28, 5.0-40, 5.0-41, B 3.4.14-2, B 3.4.19-2 and B 3.4.19-3. Enclosure 3, which includes the TS pages revised in response to the RAls, replaces Enclosure 3 of Reference 2 in its entirely. 700 First Street Hudson, Wisconsin 54016 Telephone: 715-377-3300
Document Control Desk Page 2 The additional information provided in this supplement does not impact the conclusions of the Determination of No Significant Hazards Consideration and Environmental Assessment presented in the referenced February 16, 2006 submittal as supplemented July 21, 2006. In accordance with 10 CFR 50.91, NMC is providing a copy of this letter and enclosures to the designated State Official. Summary of Commitments This letter contains no new commitments and no revisions to existing commitments. I declare under penalty of perjury that the foregoing is true and correct. Executed on 27 December 2006. Gabor Salamon Acting Director, Nuclear Licensing and Regulatory Services Nuclear Management Company, LLC Enclosures (3) cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Senior Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC State Official, Minnesota Department of Commerce
Enclosure 1 Nuclear Regulatory Commission Requests for Additional Information and Nuclear Management Company Responses The NRC Staff provided comments on revisions of Nuclear Management Company (NMC) proposed Technical Specifications (TS) and Bases in emails dated August 29, 2006, October 25,2006 and November 9,2006. The NRC Staff comments and NMC responses are provided below for each TS paragraph or Bases page on which comments were received. Page numbers refer to the page in Enclosure 2. General Comment October 25,2006, Comment 1 Any strikeoutslunderlines should be based on their current Tech Specs (not previous versions of their proposal since their previous versions are inconsequential). As currently written, it is almost impossible to read without spending an inordinate amount of time. All that matters is the currently approved TS and the current proposal. Prior versions of the proposal are immaterial. If they keep it as is, there will be a lot of effort involved in reviewing their submittal. NMC Response: The current TS and Bases were marked up to show additions and deletions incorporating Technical Specifications Task Force (TSTF) industry traveler 449 (TSTF-449) with consideration for all NRC requests for additional information (RAls) and email comments. Additions and deletions associated with previous versions have been removed. Enclosures 2 and 3 to this letter show the current proposal. TS 5.5.8.b.1, Pane 5.0-1 3 August 29,2006, Comment 2 A safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials was indicated in TS Section 5.5.8.b.1. GL 95-05 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 POB [probability of burst] of 1x10-2. As currently proposed, the flaws to which the voltage-based ARC is applied must maintain a safety factor of 1.4 against burst during design basis accidents. Since this is inconsistent with the staffs original approval (as evidenced by the probability of burst criteria), please verify that this was your intent. If this was not your intent, please discuss your plans to modify your submittal to address this Page 1 of 17
L-HU-06-037 issue. Discuss your plans to clarify your proposal, for example: "This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.8.c.2(c), a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials." NMC Response: This TS paragraph was revised to incorporate the wording as proposed by the NRC Staff and now states: This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.8.c.2(c), a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials." TS 5.5.8.b.1, Page 5.0-14 August 29,2006, Comment 1 In your proposed Structural Integrity Performance Criteria (SIPC) in TS 5.5.8.b.1, you stated the following: "For Unit 2, when tubes are left in service with predominantly axially oriented stress corrosion cracking at the tube support plate (TSP) elevations, the probability of burst (POB) under main steam line break conditions shall be maintained below 1E-02 in accordance with the requirements of NRC Generic Letter (GL) 95-05." As currently proposed, once tubes are left in service with predominantly axially oriented stress corrosion cracking at the tube support plate elevations, the probability of burst for all indications (even those that are not axially oriented stress corrosion cracking at TSP locations) is limited to 1x10-2. In addition, since NRC GL 95-05 does not contain any "requirements," the last portion of this statement is not accurate. If it was not your intent to have the 1x10-2 criteria apply to all forms of degradation, please discuss your plans to modify your submittal. Please discuss your plans to address the above. The proposed TS may be modified by using something similar to the following: For Unit 2, when alternate repair criteria discussed in Specification 5.5.8.c.2(c) are applied to axially oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG [steam generator] will burst under postulated main steam line break conditions shall be less than 1x10-2. Page 2 of 17
L-HU-06-037 Please note that your Bases may also need to be revised to clarify this issue. NMC Response: This TS paragraph was revised to incorporate the wording as proposed by the NRC Staff and now states: For Unit 2, when alternate repair criteria discussed in Specification 5.5.8.c.2(c) are applied to axially oriented outside diameter stress corrosion cracking indications at the tube support plate locations, the probability that one or more of these indications in an SG will burst under postulated main steam line break conditions shall be less than 1E-02. The Bases changes are discussed below. 5.5.8.b.2, Pane 5.0-14 August 29,2006, Comment 3 Regarding TS 5.5.8.b.2, you reference the "voltage-based repair criteria." Since this reference isn't specific, it could be misinterpreted to apply to any flaws to which a voltage-based sizing method is applied. As a result, discuss your plans to clarify your proposed TS to indicate that the "voltage-based repair criteria" that you are referring to is the one in TS 5.5.8.c.2(c). NMC Response: This TS paragraph was revised to specifically reference Specification 5.5.8.c.2(c). 5.5.8.c.2, Pane 5.0-14 August 29,2006, Comment 4 As currently written, it is not clear whether all of the criteria listed under TS 5.5.8.c.2 must be met in order to require plugging or repair. In addition, the criteria under TS 5.5.8.c.2 not only discuss the criteria for plugging and repair, but also criteria for leaving flaws in service. As a result, please discuss your plans to modify your submittal to address this issue. For example: "Unit 2 steam generator tubes found by inservice inspection to contain flaws shall be dispositioned as follows:"
L-HU-06-037 NMC Response: This TS paragraph was revised to incorporate the wording as proposed by the NRC Staff and now states, "Unit 2 steam generator tubes found by inservice inspection to contain flaws shall be dispositioned as follows:". 5.5.8.c.2(a)(l), Page 5.0-20 August 29,2006, Comment 5 It appears that TS 5.5.8.~.2(a)(l) and TS 5.5.8.c.2(a)(2) are intended to address the repair criteria for the non-sleeved and sleeved region of the tube, respectively. In your current proposal (and TSTF-449), a "tube" is considered to include the tube wall and any repairs to it. As a result, it would appear that there are two different set of repair limits for the sleeves (since TS 5.5.8.c.2(a)(l) and TS 5.5.8.c.2(a)(2) apply to the sleeve). Please discuss your plans to clarify that TS 5.5.8.c.2(a)(l) addresses the non-sleeved region of the tube and TS 5.5.8.c.2(a)(2) addresses the sleeved region of the tube. October 25,2006, Comment 2 In TS 5.5.8.c.2(a)(l), they should broaden the exception to include the F*/EF* criteria (i.e., except if permitted to remain i n service through application of the alternate tube repair criteria discussed in Specification 5.5.8.c.2(b) or Specification 5.5.8.c.2(c). NMC Response: This TS paragraph was revised to apply only to a flaw in a non-sleeved region of the tube and the exception recommended in the October 25,2006, Comment 2 was incorporated. 5.5.8.c.2(a)(2), Page 5.0-20 August 29,2006, Comment 5 It appears that TS 5.5.8.c,2(a)(l) and TS 5.5.8.c.2(a)(2) are intended to address the repair criteria for the non-sleeved and sleeved region of the tube, respectively. In your current proposal (and TSTF-449), a "tube" is considered to include the tube wall and any repairs to it. As a result, it would appear that there are two different set of repair limits for the sleeves (since TS 5.5.8.c,2(a)(l) and TS 5.5.8.c.2(a)(2) apply to the sleeve). Please discuss your plans to clarify that TS 5.5.8.c.2(a)(l) addresses the non-sleeved region of the tube and TS 5.5.8.c.2(a)(2) addresses the sleeved region of the tube. Page 4 of 17
L-HU-06-037 August 29,2006, Comment 6 In proposed TS 5.5.8.c.2(a)(2), you indicated that the repair criteria for the original tube wall in the sleeve to tube joint is 25-percent of the nominal sleeve wall thickness. This does not appear to be consistent with your current technical specifications (and it probably is not consistent with the design and licensing basis for the sleeves). The staff believes that you intended to indicate that the repair criteria for the sleeve is 25-percent of the sleeve wall thickness and that the repair criteria for the parent tube at the sleeve-to-tube joint is to plug on detection. Please discuss your plans to modify your proposal to address this issue. In addition, as currently written, proposed TS 5.5.8.c.2(a)(2) would permit tubes to be either plugged or repaired in the event that flaws exceeded the repair criteria. Please discuss your plans to indicate that flaws that exceed these repair limits must be plugged. NMC Response: The TS paragraph was revised to apply only to a flaw in a sleeved region of the tube and require tube plugging when the criterion is exceeded. A new TS paragraph 5.5.8.c.2(a)(3) was added to require plugging of tubes with a flaw in a sleeve to tube joint. August 29,2006, Comment 6 In proposed TS 5.5.8.c.2(a)(2), you indicated that the repair criteria for the original tube wall in the sleeve to tube joint is 25-percent of the nominal sleeve wall thickness. This does not appear to be consistent with your current technical specifications (and it probably is not consistent with the design and licensing basis for the sleeves). The staff believes that you intended to indicate that the repair criteria for the sleeve is 25-percent of the sleeve wall thickness and that the repair criteria for the parent tube at the sleeve-to-tube joint is to plug on detection. Please discuss your plans to modify your proposal to address this issue. In addition, as currently written, proposed TS 5.5.8.c.2(a)(2) would permit tubes to be either plugged or repaired in the event that flaws exceeded the repair criteria. Please discuss your plans to indicate that flaws that exceed these repair limits must be plugged. Page 5 of 17
L-HU-06-037 October 25,2006, Comment 3 TS 5.5.8.c.2(a)(3) should be clarified to indicate that "Tubes with a flaw in a sleeve to tube joint that occurs in the original tube wall of the joint shall be plugged." (to avoid potential overlap with the prior requirement). NMC Response: This TS paragraph was added to require plugging of tubes with a flaw in a sleeve to tube joint and the phrasing recommended in the October 25, 2006, Comment 3 was incorporated. 5.5.8.c.2(b), Pane 5.0-20 August 29,2006, Comment 7 In proposed TS 5.5.8.c.2(b), it would appear that the following phrase is not needed since it is also contained in proposed TS 5.5.8.c.2(b)(l) and (2): "Flaws may be left in service when they are located below F* or EF* [region] defined below:." Please discuss your plans to remove this phrase. August 29,2006, Comment 15 In your July 21,2006 response to question 3, you stated (see item
- 2) that the F* and EF* criteria could be applied to the cold-leg side of the tubesheet. At the time the F* and EF* criteria were approved, your technical specification only addressed the hot-leg portion of the tubesheet (i.e., no inspections were required by the technical specifications in the cold-leg). At the time of these F* and EF* proposals, no modifications were made to the technical specifications to require cold-leg inspections. As a result, the staff reviewed your proposal to incorporate technical specification inspection and repair criteria for the hot-leg. As a result of the above, discuss your plans to submit for review and approval, the structural and leakage integrity analysis for application of the F*
and EF* criteria to the cold-leg or alternatively discuss your plans to clarify that the F* and EF* criteria apply to the hot-leg. October 25,2006, Comment 4 Regarding the F*/EF* criteria, reference should be made to TS 5.5.8.c.2(a)(l). That is, "....alternative to the depth based criteria in Specification 5.5.8.~.2(a)(l) (since it would be inappropriate to apply this to the depth based criteria of sleeves). Page 6 of 17
L-HU-06-037 October 25,2006, Comment 5 It would seem that it should be clear that the F* criterion does not apply to tubes that have a sleeve installed below the uppermost hardroll transition. October 25,2006, Comment 6 For the voltage based repair criteria, I have similar comments as made above regarding the F* criterion. November 9,2006, Comment 9 We still have the issue with respect to the F*/EF* criteria and the cold leg. NMC Response: This TS paragraph was revised to remove the sentence, ""Flaws may be left in service when they are located below F* or EF* [region] defined below: . ." and specifically reference Specification 5.5.8.c.2(a)(l). NMC agrees that the F* criterion and voltage based repair criteria do not apply to tubes that have a sleeve installed below the uppermost hardroll transition. No TS changes were made to address these comments. NMC has added, "to the hot-leg of the tubesheet" in this paragraph to restrict the use of the F* and EF* criteria to the hot-leg. 5.5.8.c.2(b)(l), Page 5.0-21 and 5.5.8.cn2(b)(2),Page 5.0-21 August 29,2006, Comment 8 In several instances, the term "defect" is used in your proposed TS (e.g., 5.5.8.c.2(b)(l), proposed TS 5.5.8.c.2(b)(2), and proposed TS 5.6.7.a.10). Since a "defect" is not defined in your proposed TS, please discuss your plans to replace this term with "flaw" which is the term used in TSTF-449. In addition, the term "degradation" is used in your proposed TS (e.g., 5.5.8.c.2(c)(l) and 5.5.8.c.2(~)(2)). Since "degradation" is not defined in your proposed TS, please discuss your plans to replace this term with "flaw" which is the term used in TSTF-449. November 9,2006, Comment I In 5.5.8.c.2.b.l and 5.5.8.c.2.b.2, it appears that the second from the last sentence should be modified (This 1.07-inch span (not including eddy current uncertainty) is referred to as the F* region.) I believe this sentence should be modified to Page 7 of 17
L-HU-06-037 indicate that "This 1.07-inch span (when increased for eddy current uncertainty) is referred to as the F* region." The corresponding change should also be made to the EF* section. The reason for the change is that the F* region definition is used to indicate that all tubes with flaws in this region should be plugged or repaired. As currently written, one could interpret the sentence as the F* region does not include eddy current uncertainty (which is not the correct interpretation). NMC Response: These TS paragraphs were revised to replace "defects" with "flaws" as proposed by the NRC Staff. The parenthetical statements following the last mention of the 1.07-inch span in TS 5.5.8.c.2(b)(l) and the last mention of the 1.67-inch span in TS 5.5.8.c.2(b)(2) were revised to state, "increased for measurement uncertainty" as agreed to in a phone call with the NRC Staff on November 21, 2006. Revisions to TS 5.5.8.c.2(~)(1),TS 5.5.8.c.2(~)(2)and TS 5.6.7.a.10 are discussed below. 5.5.8.c.2(c), Pane 5.0-21 November 9,2006, Comment 2 In 5.5.8.c.2.c, they should add "....as an alternative to the depth based criteria in Specification 5.5.8.c.2(a)(l)." This will make it similar to the F* criterion writeup (but more appropriately it clarifies that these alternate repair criteria can only be applied to the non-sleeved portion of the tube). NMC Response: This TS paragraph was revised to incorporate the wording as proposed by the NRC Staff and now states, ". . .as an alternative to the depth based criteria in Specification 5.5.8.c.2(a)(l)11. 5.5.8.c.2(c)(l), Page 5.0-21 and 5.5.8.c.2(~)(2), Paqes 5.0-21 and 5.0-22 August 29,2006, Comment 8 In several instances, the term "defect" is used in your proposed TS (e.g., 5.5.8.c.2(b)(l), proposed TS 5.5.8.c.2(b)(2), and proposed TS 5.6.7.a.10). Since a "defect" is not defined in your proposed TS, please discuss your plans to replace this term with "flaw" which is the term used in TSTF-449. In addition, the term "degradation" is used in your proposed TS (e.g., 5.5.8.c.2(~)(1) and 5.5.8.c.2(~)(2)). Page 8 of 17
L-H U-06-037 Since "degradation" is not defined in your proposed TS, please discuss your plans to replace this term with "flaw" which is the term used in TSTF-449. NMC Response: These TS paragraphs were revised to replace "degradation1'with "indication" since bobbin voltages identify "indications" rather than "flaws1'. 5.5.8.d, Pane 5.0-26 August 29,2006, Comment 9 Please discuss your plans to indicate in TS 5.5.8.d that: "In tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection." NMC Response: This TS paragraph was revised to incorporate the wording as proposed by the NRC Staff. 5.5.8.d.3(a), Pane 5.0-27 August 29,2006, Comment 10 In proposed TS 5.5.8.d.3(a), you indicate that the region of the tube below the F* and EF* regions may be excluded from the inspection requirements. In addition, in your response to question 4c in your July 21,2006 letter (ML062370052), you indicate that full depth tubesheet sleeves are installed at the lower end of the parent tube (presumably this is near the tube-to-tubesheet weld). Since this latter region is below the F* and EF* region, it would appear that a tube in which a full depth tubesheet sleeve is installed may not require an inspection near the lower end of the sleeve (depending on exactly where the sleeve is installed with respect to the F* and EF* region). As a result, please discuss your plans to modify your proposal to ensure that full depth tubesheet sleeves require an inspection. August 29,2006, Comment 11 In proposed TS 5.5.8.d.3(a), you reference a "refueling outage inspection." Under the proposed TS, inspections need not be performed during a refueling outage. They only need to be performed at intervals not to exceed 24 effective full power months or one operating interval between refueling outages (whichever is less). As a result, if you were to elect to perform inspections at times other than
L-HU-06-037 refueling outages, the F* and EF* region may not be inspected for multiple cycles. Since this is inconsistent with your current requirements (and the designllicensing basis), discuss your plans to modify your submittal to indicate that the "F* and EF* tubes" will be inspected in the F* and EF* regions every 24 effective full power months or one refueling outage (whichever is less). A similar comment applies to proposed TS 5.5.8.d.3(c) which references inspections during refueling outages. October 25,2006, Comment 7 In 5.5.8.d.3.a, the term "periodic" is introduced. Since this is confusing it should be rewritten (e.g., one may interpret this as the 60 month periodic inspection and this would be inappropriate). I would suggest terminology such as "every 24 EFPMs or one refueling outage (whichever is less)." A similar comment applies to other uses of "periodic" November 9,2006, Comment 3 In 5.5.8.d.3.a, the last sentence is awkward. Iwould suggest the following: The region of these tubes below the F* and EF* regions do not need to be inspected unless there is a sleeve (or portion of a sleeve) that extends below the F* or EF* region. NMC Response: The NRC comments have been resolved through adoption of the parenthetical phrase, "every 24 effective full power months (EFPM) or one refueling outage (whichever is less)", in the first sentence as suggested by the NRC Staff in the October 25, 2006 Comment 7 and revision of the last sentence as suggested in the November 9, 2006 Comment 3. Resolution of these comments as applicable to other TS paragraphs is discussed below. 5.5.8.d.3(b), Pane 5.0-27 August 29,2006, Comment 12 In proposed TS 5.5.8.d.3(b) and (c), you refer to the repair criteria discussed in proposed TS 5.5.8.c.2(c) using different terminology. This can cause confusion on what is being referred to (since neither of these sections match the "title" in 5.5.8.c.2(c). As a result, please discuss your plans to modify these two sections to simply reference the "alternate repair criteria discussed in TS 5.5.8.c.2(~)." A similar comment applies to proposed TS 5.6.7.b. Page 10 of 17
L-HU-06-037 NMC Response: This TS paragraph was revised to incorporate the wording as proposed by the NRC Staff. August 29,2006, Comment 11 In proposed TS 5.5.8.d.3(a), you reference a "refueling outage inspection." Under the proposed TS, inspections need not be performed during a refueling outage. They only need to be performed at intervals not to exceed 24 effective full power months or one operating interval between refueling outages (whichever is less). As a result, if you were to elect to perform inspections at times other than refueling outages, the F* and EF* region may not be inspected for multiple cycles. Since this is inconsistent with your current requirements (and the designllicensing basis), discuss your plans to modify your submittal to indicate that the "F* and EF* tubes" will be inspected in the F* and EF* regions every 24 effective full power months or one refueling outage (whichever is less). A similar comment applies to proposed TS 5.5.8.d.3(c) which references inspections during refueling outages. August 29,2006, Comment 12 In proposed TS 5.5.8.d.3(b) and (c), you refer to the repair criteria discussed in proposed TS 5.5.8.c.2(c) using different terminology. This can cause confusion on what is being referred to (since neither of these sections match the "title" in 5.5.8.c.2(c). As a result, please discuss your plans to modify these two sections to simply reference the "alternate repair criteria discussed in TS 5.5.8.c.2(c)." A similar comment applies to proposed TS 5.6.7.b. October 25,2006, Comment 7 In 5.5.8.d.3.a, the term "periodic" is introduced. Since this is confusing it should be rewritten (e.g., one may interpret this as the 60 month periodic inspection and this would be inappropriate). Iwould suggest terminology such as "every 24 EFPMs or one refueling outage (whichever is less)." A similar comment applies to other uses of "periodic" November 9,2006, Comment 4 In 5.5.8.d.3.~~they should confirm that the Spec referenced is 5.5.8.c.2(c) since I could not read portions of the spec in the hard copy that I have. Page 11 of 17
L-H U-06-037 NMC Response: The NRC comments have been resolved through adoption of the phrase, "every 24 effective full power months (EFPM) or one refueling outage (whichever is less)", as suggested by the NRC Staff in the October 25, 2006 Comment 7 and referencing the alternate repair criteria as suggested by the NRC Staff in the August 29, 2006 Comment
- 12. The reference to Specification 5.5.8.c.2(c) is correct.
5.5.8.d.3(d) (Not included in the current proposed TS) August 29,2006, Comment 10 In proposed TS 5.5.8.d.3(a), you indicate that the region of the tube below the F* and EF* regions may be excluded from the inspection requirements. In addition, in your response to question 4c in your July 21,2006 letter (ML062370052), you indicate that full depth tubesheet sleeves are installed at the lower end of the parent tube (presumably this is near the tube-to-tubesheet weld). Since this latter region is below the F* and EF* region, it would appear that a tube in which a full depth tubesheet sleeve is installed may not require an inspection near the lower end of the sleeve (depending on exactly where the sleeve is installed with respect to the F* and EF* region). As a result, please discuss your plans to modify your proposal to ensure that full depth tubesheet sleeves require an inspection. October 25,2006, Comment 8 The frequency should be added to 5.5.8.d.3.d (i.e., every 24 EFPM or 1 RFO). The reference to sleeving is awkward (i.e., inspect 100% of the insewice tubes in the non-sleeved tubesheet region) since the tubesheet isn't sleeved. The easiest fix would be to delete non-sleeved. Alternatively wording such as the following should be considered, "For tubes with no portion of the sleeve within the [hot leg] tubesheet region, inspect 100% of the inservice tubes in the [hot-leg] tubesheet region, . . . when the F* or EF* methodology has been implemented." November 9,2006, Comment 5 It is not clear why they deleted 5.5.8.d.3.d. Our preference is that they retain it (as modified based on our previous comments). NMC Response: In response to the August 29, 2006 Comment 10, NMC included additional requirements in a draft proposed TS based on another plant's submittal. After further review, NMC realized that the additional requirements were beyond the Prairie Island Nuclear Generating Plant (PINGP) licensing basis. Current TS 5.5.8.b.3 states: Page 12 of 17
L-HU-06-037 In addition to the sample required in Specification 5.5.8.b.2(a) through (c), all tubes which have had the F* or EF* criteria applied will be inspected in the F* and EF* regions of the roll expanded region. The region of these tubes below the F* and EF* regions may be excluded from the requirements of Specification 5.5.8.b.2(a). These current TS requirements are embodied in the requirements of TS 5.5.8.d.3(a) proposed in this supplement and thus this paragraph is not included in this supplement. August 29,2006, Comment 13 In proposed TS 5.5.8.f.2, you indicate that hardroll expanding portions of tubes in the tubesheet is an acceptable tube repair method. Since a tube may includes a sleeve, please discuss your plans to clarify that this repair criteria is only applicable to tubes that do not have sleeves installed in the tubesheet region. For example, "Hardroll expanding non-sleeved portions of tubes in the tubesheet in order to apply the F* and EF* criteria." NMC Response: This TS paragraph was revised to incorporate the wording as proposed by the NRC Staff. 5.6.7.a.10, Page 5.0-40 August 29,2006, Comment 8 In several instances, the term "defect" is used in your proposed TS (e.g., 5.5.8.c.2(b)(l), proposed TS 5.5.8.c.2(b)(2), and proposed TS 5.6.7.a.10). Since a "defect" is not defined in your proposed TS, please discuss your plans to replace this term with "flaw" which is the term used in TSTF-449. In addition, the term "degradation" is used in your proposed TS (e.g., 5.5.8.c.2(c)(l) and 5.5.8.c.2(~)(2)). Since "degradation" is not defined in your proposed TS, please discuss your plans to replace this term with "flaw" which is the term used in TSTF-449. November 9, Comment 6 In 5.6.7.a.10, the specification referenced should be 5.5.8.d (not 5.5.8.d.3(a)). Our suggestion would make it consistent with their current spec. Their proposal limits their current reporting requirement. Page 13 of 17
L-HU-06-037 NMC Response: This paragraph was revised to refer to "flaws1'as proposed by the NRC Staff. Since F* and EF* only apply to Unit 2, the specification referenced is TS 5.5.8.d.3. 5.6.7. b, Page 5.0-40 August 29,2006, Comment 12 In proposed TS 5.5.8.d.3(b) and (c), you refer to the repair criteria discussed in proposed TS 5.5.8.c.2(c) using different terminology. This can cause confusion on what is being referred to (since neither of these sections match the "title" in 5.5.8.c.2(c). As a result, please discuss your plans to modify these two sections to simply reference the "alternate repair criteria discussed in TS 5.5.8.c.2(~)." A similar comment applies to proposed TS 5.6.7.b. October 25,2006, Comment 9 On page 5.0-40, requirement "b ...",it does not appear that "to tube support plate intersections" is needed. In fact, maybe more appropriate wording should be, "When the alternate repair criteria discussed in...... are implemented, notify....." NMC Response: This TS paragraph has been revised by specifically referencing "alternate repair criteria discussed in TS 5.5.8.c.2(c)l8and deleting "to tube support plate intersections" as suggested by the NRC Staff. 5.6.7.b.4, Page 5.0-41 August 29,2006, Comment 14 Regarding proposed TS 5.6.7.b.4, you indicated that removing this reporting requirement would constitute a change in your licensing basis (refer to your response to question 2 in the July 21, 2006 letter). The staff notes that by incorporating the 1x10-2 probability of burst criteria into TS 5.5.8.b.1, you will not be able to operate under the condition where the burst probability exceeds 10-2. As a result, providing a safety assessment is not needed. As a result, the reporting requirement is not needed. The staff also notes that you are required per 10 CFR 50.73 to report if the performance criteria are not maintained. As a result of the above, discuss your plans to remove the subject reporting requirement. Page 14 of 17
L-HU-06-037 NMC Response: Current TS requirement 5.6.7.5.e has been deleted as recommended by the NRC Staff. Bases B 3.4.14, Page 3.4.14-2 August 29,2006, Comment 3 Regarding TS 5.5.8.b.2, you reference the "voltage-based repair criteria." Since this reference isn't specific, it could be misinterpreted to apply to any flaws to which a voltage-based sizing method is applied. As a result, discuss your plans to clarify your proposed TS to indicate that the "voltage-based repair criteria" that you are referring to is the one in TS 5.5.8.c.2(c). November 9,2006, Comment 7 In the first paragraph on B 3.4.14-2, they do not include the last sentence that the TSTF indicated should be included. Namely, "The LC0 requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis." Is there a reason for this? NMC Response: The first and last paragraphs on this Bases page were revised to reference TS 5.5.8.c.2(c) as suggested in the August 29, 2006 Comment 3. The TSTF-449 sentence was restored to this page in accordance with the November 9, 2006 Comment 7. Bases B 3.4.1 9, Page B 3.4.1 9-2 August 29,2006, Comment 1 In your proposed Structural Integrity Performance Criteria (SIPC) in Technical Specification (TS) 5.5.8.b.1, you stated the following: "For Unit 2, when tubes are left in service with predominantly axially oriented stress corrosion cracking at the tube support plate (TSP) elevations, the probability of burst (POB) under main steam line break conditions shall be maintained below 1E-02 in accordance with the requirements of NRC Generic Letter (GL) 95-05." As currently proposed, once tubes are left in service with predominantly axially oriented stress corrosion cracking at the tube support plate elevations, the probability of burst for all indications (even those that are not axially oriented stress corrosion cracking at TSP locations) is limited to 1x10-2. In addition, since NRC GL 95-05 does not contain any "requirements," the last portion of this statement is not accurate. If it Page 15 of 17
L-HU-06-037 was not your intent to have the 1x10-2 criteria apply to all forms of degradation, please discuss your plans to modify your submittal. Please discuss your plans to address the above. The proposed TS may be modified by using something similar to the following: For Unit 2, when alternate repair criteria discussed in Specification 5.5.8.c.2(c) are applied to axially oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than 1x10-2. Please note that your Bases may also need to be revised to clarify this issue. August 29,2006, Comment 3 Regarding TS 5.5.8.b.2, you reference the "voltage-based repair criteria." Since this reference isn't specific, it could be misinterpreted to apply to any flaws to which a voltage-based sizing method is applied. As a result, discuss your plans to clarify your proposed TS to indicate that the "voltage-based repair criteria" that you are referring to is the one in TS 5.5.8.c.2(c). October 25,2006, Comment 10 On page B 3.4.19-2, there appears to be a typo "thes" should be "these". NMC Response: The second paragraph of the Applicable Safety Analyses discussion was revised to specifically reference Specification 5.5.8.c.2(c) as suggested by the NRC Staff (August 29, 2006, Comment 3) and the typographical error was corrected. Discussion about axially oriented outside diameter stress corrosion cracking indications similar to that proposed by the NRC Staff (August 29, 2006, Comment I) was also included in this paragraph. Bases B 3.4.1 9, Pane B 3.4.1 9-3 August 16,2006, Comment 16 In the Limiting Condition for Operation section of B 3.4.19, you indicate that the F* and EF* distances are not considered part of the tube. Since these distances are no longer defined in your proposed TS, please discuss your plans to modify this phrase to indicate that the region of tube below the F* and EF* regions is not considered part of the tube. In addition, discuss your plans to indicate that the Page 16 of 17
L-HU-06-037 parent tube (original tube wall) between sleeve joints is also not considered part of the tube. October 25,2006, Comment 11 On page B 3.4.19-3, the wording will need to be clarified since a sleeve installed below the F* and EF* region is still part of the tube (i.e., when a sleeve is installed, there is still an F*IEF* region it's just no longer part of the pressure boundary). November 9,2006, Comment 8 In the 3rd paragraph in the LC0 section on page B 3.4.19-3, they should remove the "(sleeves)" qualifier on repairs since it is not needed. In addition, it is not clear that the last sentence is complete. We would recommend something like the following: The tube-to-tubesheet weld is not considered part of the tube, nor is the region of the tube below the F* and EF* region (provided no sleeve extends below the F* and EF* region in which case the sleeve is part of the tube), nor the portion of the original tube wall between the sleeve joints. NMC Response: The third paragraph in the Limiting Condition for Operation discussion on this page was revised by removing "(sleeves)" which was included in a draft version. The other issues in the comments applicable to this paragraph were resolved by the addition of a parenthetical clause "(except as noted below)" and an additional sentence to which the NRC Staff agreed in a phone call on November 21,2006. Page 17 of 17
ENCLOSURE 2 Proposed Technical Specification and Bases Pages (markup) Prairie Island Nuclear Generating Plant Units I and 2 Technical Specification Pages Bases pages 46 pages follow
Definitions 1.1 1.1 Definitions (continued) E -AVERAGE E shall be the average (weighted in proportion to the concentration D I S r n R A JION of each radionuclide in the reactor coolant at the time of sampling) ENERGY of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant. LEAKAGE LEAKAGE from the Reactor Coolant System (RCS) shall be:
- a. Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
- 3. RCS LEAKAGE through a steam generator @-+ot the Secondary System (primary to secondary LEAKAGE);
- b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;
- c. Pressure Boundary LEAKAGE LEAKAGE (except prin~ar~_t:c.ss.condarv %LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
Prairie Island Unit 1 - Amendment No. 43-8 Units 1 and 2 1.1-3 Unit 2 - Amendment No. 449
RCS Operational LEAKAGE 3.4.14 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. RCS identified 6 hours LEAKAGE not within limit for reasons other AND than pressure boundary LEAKAGE................. or prima~v A to C.2.1 Reduce LEAKAGE to
................. 14 hours secondary LEAKAGE. within limits.
OR C.2.2 Be in MODE 5. 44 hours D. Pressure boundary D. 1 Be in MODE 3. 6 hours LEAKAGE exists.
\No D.2 Be in MODE 5. 36 hours Primar~tosec~.nda~:.SG LEAKAGEnot within limit.
Prairie Island Unit 1 - Amendment No. 44% Units 1 and 2 3.4.14-2 Unit 2 - Amendment No. 4-49
RCS Operational LEAKAGE 3.4.14 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ............................ NOTES. ......................... L N o t required to be
..... until 12 hours after establishment of steady state operation.
- 2. Not applicable to primary to secondarv LEAKAGE.
Verify RCS operational I,EA.Ke?G.Eledagewithin limits by performance of RCS water inventory 24 hours balance. SR 3.4.14.2 ........................... NOTE........................... Not required to be performed until 12 hours after establishment of steady state operation. Verify f b ' i J L m p r i r n a - y to secondary 1,EAKAGII is < 150 gallons per day through any one SG.. Prairie Island Unit 1 - Amendment No. 43-8 Units 1 and 2 Unit 2 - Amendment No. 149
SG Tube Integrity 3.4 -- REACTO_R_COOLANTSYSTEM (RCS) 3.4.1-9 Stcain --Ggnerator_LSIr;iTiibe - Integrity LC0 - -- 3.4.19 S(; tube integrity shall be maintained.
--- _ AND
- - - A11 SG tubes satisbing the t u b c r ~ g i criteria r shall
- be plugged or repaired in accordance with the Steanl Generator Program.
AI'I'LlCABILI'l'Y: MODES 1, 2. 3 : and 4. N O'j'l-{.................................................. &grate Condition entry is allowed for each SG tube. CONDITION REQUIRED ACTION COMPLETION satisfying the tube the affected tube(s) is rcpair critcria and not plugged or remired in refucl ing oukiige -or S(3 ziccordance with the Steam Generator tube(s) in accordmcc MODE 4 following Prairie Island _ - - --- - - - - tJnit 1 - Amendment- No. Units 1 and 2 3.4.19-1 _ Unit 2 - Amendment No.
SC; Tube Integrity
................................................ ............ -. .................................... - . 3.4.19 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME H . K ~ u i r e dAction and 6 hours associated C o i n p l e m Time of CorditLon A not cnct.
-36 hours - SG tube integrity not maintained.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR.
. Veri fs.S G tube -- int-eg,r~~y.~-n:c!.c.c.~.rda.nc_e. ..~xith...th~I n accordance Steg.CJ.g:n:r.&or r).l:Oigg.mmmf. yy&h the Stcaln CJ-cnerator l'rogran~
Prior to entering . MODE4 a.c.c.~rdance
.. ~~
... i t h SteamG.e.ncr.ator
.... . the.. -followin&~..sG -
tube.i.nsa~ction Prairie Isl~ind - -- - - - - ----- - ----- lJnit 1 - Amendment No. Ilnits I and 2- 3.4.19-2 Unit 2 - Amendment No.
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 Steam Generator (SG) -Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition. the Steam C)_gneratorh o g r a ~ nshdl include thq lbllo\yi~gpr~visic~ns: a ...Provi.si.ons
... %. . . ..... .. ~ Ir_...c I .on.d.itI~!l ....m.o.n.itor.in_g ....~.s.se~.s~n.e.nt.s..~
.....Condition
!ggnto.r.i.ngasses.sn~.e.nt..~n.ca..ns...a.n...evalu ... d . the '.'as....fou.n.cl'.'
atio.n.= . conditionof-th.etub.i.ng...withres~ecttothe..~eyfo~~.~ancc=..cr.i.t . eri.a-fo1: structural integrity and accident induced leakage. 'I'he "as found condition refers to the condition of the tubing during an SG inspection outage, as determined from th.e inservice inspection results or by other means. prior to the dugring oaepair of tubes.-Condition monitoring a.ss.e.s.sments...sha!.I.be....
-- ~~~.n.du-~te.d ...d4i.nglilie.aad~ . 0.uta~.e...~.u1:~g~whi.ch ...th-e S.G...t.ul~:s..areins~ect.ed.~ ....p1.ug1?e.d.~ ....or...r.e.p~i.r.ed_.t.~ ..w-~!.fi.r.n~...thtt.t_h.e performance criteria are being met.
- b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained bv meeting the perlbrmance criteria for tube structural integrity. accident induced leakage. and operational LEAKAGE.
................. .... .Structural...i.nte~~.it~~..~_e_rf~~1:n~;~1~?.c.e_.c~riter.i.o~~.~ . All i.n-.s..... .en!ice. stern
&ie-cs@or..tube_s_.s h11.1 ....re.tain....struc.t~~a! ~~_te~i.t~~.~e.ver...the
.....i .....@ !!..range of normal operating conditions (including startup. operation in the power range. hot standby, and cool down and all anticipated transi~~lts. .i~n~luded....ir?._the....d.e.s..Ig,n...s.p.e.c.i.tZc.a~:IonI.an_d de~i~gnhsis accidem. Thkinc!udes~etai.r!i.nga...s.ak_t:~~ .. f~:;ic~nroL3,Q.aggainst 12.~xis.t. under...n~r~m:al.~t:gd.y.::st@
... fuVpnwer ~ ~ . p . e g ~ i o n ~ . ~ i m a w - t . ~ ~
s.c.c~.nd.a.~..nressu.r.~ ....d1.f.fe~.e.nt.i~l~~~n~~:.e~c~ep.t~~.r .. flg~sS:~.dddreess.ed throuh ...a~_n!iLc_c2tion. ...~.f.the....a!.te.~.ate...r~~.a.i r...crite.rl.aaadiscussedic Specification 5.5.8.c.2(c). a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apai-t from the above requirements. additional .-
. . .loading; conditions associated with thewdesign basis accidents, or ...
ccrmbi.natic?.nof ..~~.c~id..ent.s....i ... .r?.....a:.c.~~rd.an~.e....~it1_!~~~t1~.e....d.e.s~n . and
!.i.ce!lsi.ng.ba.siissZ
. - shall a!..s.o....b..e....e.1~.al~uate_~ta.dete.r.mi.neif
.... ... the a_sl.s_~ci.ated. ...!-o_a_ s!_s.... con~~~.bu_te.~s~.~n.i~~.ca.nt!.~...t.~...b~~rst. -(~,_r-.c.
o!.!.a~s.e.~. ... the assessment of tube integrity. those loads that do significantly Prairie Island Unit 1 - Amendment No. 44% Units 1 and 2 5 .O- 13 Unit 2 - Amendment No. 449
Programs and Manuals 5.5 affect...burst.oro!.!.a~s$.. la!..!.....bec!eter!.n.i.ned...a!!.d . .a s s s e d i n . combination
........... with the loads due to *pressure
.................................................................................................................... with a safety..................................................
.............................................................................................. factor of 1.2 on the combined primarv loads and 1.0 on axial secondary loads. For Unit 2. when alternate repair criteria discussed in Sbgcification5,5.8.c.2(c) are applied to axially oriented outside diameter . ...- stress corrosion c r a c k i ~..i .nd.i.~atio.ns
?~ .....at ...t.!~.e...t.~b~t:..~s.~~.~.pc~ .rt p1.at.e...!oc;atl.ons.,.-.t!1.e ~.rirb..ab..i! .... . . i ~ ~ . . . ~ orrrrr!l~)rree ~ a t . ~ .....o
.... ~fn.th.es.e. .~
indications in. anS(;...will burg-.underaos!u!a.t_eP_.._rrl.ain
... .. .... ....swcn 1.i.n.e ...
break..conditlons sha!.! ....belessthan .IE:02,
- 2. Accident induced leakage performance criterion: 'l'he primary to secondary accident induced leakage rate for any design basis accident. other than a SG tube rupture, shall not exceed the
!.e.zik.aaerat.e.a.~~.ume.d.~i~nthe~a.~c~~.d.e.~t..~zi.na!.~.s.~~..~~~~
. ... total leakage.......rate - .. for all SGs and leaka~e - rate for an individual SG. ........
For Unit 1. leakage is not to exceed 1 mrn per
....................................................................................................................................................... SG. -...................................................
......................... For Unit 2, leakage from all sources, excluding the leakage attributed to the degradation associated with implementation of the voltage-based re.~~~rcr~i~.~~iadisc~~s~d~inSp.e~c.ifi~~ti.~~.~ . 5~..5~.S.~..~..~.2~Cc.L..~...n~~1;~
ex:e.gd .::1._gp-g::p:r:.-S.G. ,
- 3. The operational
- .................................................... I.EAKAGE~erformance
* ..-. ................................................................................ .-. criterion is specified .-......................... in I.,CO 3.4.14. "RCS ...... ........-.--
e Operational --- T..,EAKAGI.<."
- c. Provisions for SG tube repair criteria:
!.............. .IJn.it l.... s t . e a ~ ~ ~ n . ~ r . . a t ~..fo_u.ndby
.... ~ r . . . t u b....i.n.s..er~.~..~~e..m_n_s~..e.~ti.o.n~
.e~
comtai!~...fla.~s~~ltb...a...d~pth.~ee~.ur1...tto ....or....e.x.~.e.ed.i!~g ..40% of..the ... notnina!. t~rb.e...~~ia!! ...th~.c~n_e.ss..~h.a!.!~ .beeee~!~r~g;ed.! 2 Unit 2 steam generator tubes found by inservice inspection to contain flaws shall be dispositioned as follows: Prairie Island Unit 1 - Amendment No. 458 Units 1 and 2 5.0-14 Unit 2 - Amendment No. 4 4
Programs and Manuals Prairie Island Unit 1 - Amendment No. W Units 1 and 2 5.0-15 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 Prairie Island Unit 1 - Amendment No. 44% Units 1 and 2 5.0-16 Unit 2 - Amendment No. 149
Programs and Manuals 5.5 Prairie Island Unit 1 - Amendment No. 458 Units 1 and 2 5.0-17 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 1 AVT C
.. .. . . C Prairie Island Unit 1 - Amendment No. 44-8 Units 1 and 2 5.0-18 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 Prairie Island Unit 1 - Amendment No. 458 Units 1 and 2 5.0-19 Unit 2 - Amendment No. 4-49
Programs and Manuals (13 Re-wr L : t ; r l l r. f e c -5, . 1.. ...... a Tubes . found by inservice inwection containing a flaw h non-sleeved region with a depth equal to or exceeding 50% of the nominal tube wall thickness shall be plug-z@
.or repaired except if pe-mitted to remain in seryke t h i g h ap~licatiimof the alternate tube repair criteria disc~lssed-in ...S.pe.~.i.fi.cdti,.$~n.~5~~5:~8~~~21b~. .c?r..inS.p.e.c~f~.at~.on
. . If significant general tube thinning occurs, this . criterionisreduced to 40% wall penetration.
-f.2)... .......T.ub.c.s
....f0.u.
~!.cl~~b.y....
a , . inseLy:g.e.lns~-e-ctii~n:i~..r?ta.~_.n~~~.g..a
....fl.2i.w eht& - pressure boundary region of any sleeve exceedingis 25% of the nominal sleeve wall thickness shall be plug-zed.
(3) 'rubes with a flaw in a sleeve to tube ioint that occurs in
@e,riginal tube wall of the ioint shall be plugged.
[b) The following F* or EF* ..Ale.maL~.R~egair..Clriteriaan?ayYhe gppl:i:g.d..to.. the I;.~t,l.d.g,of'the ffu~~.~.~~~~~teee~.ssssaanaltem to..tls.e depth:..bs.s.ii:d.. grutg!.~.: ~.n...nSpgc_ljrca_t~.~.nn 5...5.,8,.cz2(..aj.C_!)_ Prairie Island Unit 1 - Amendment No. 4% Units 1 and 2 5 .O-20 Unit 2 - Amendment No. 449
Programs and Manuals 5.5
.. . 1 F* Criterion: If the bottom (1 _._ ...................of the .-.._ u ~ +._~
.... e
......_r n ~hardroll ost transition
. ........ - in the tubesheet is below the rnid~lane of the-.
tubesheet, then all flaws located below 1.07 inches from the bottom of this up~ermosthardroll transition (not h~c.ludi.n_g.-eddy.
. . current_u~~ce~a~ntylma~~b.e...a_!!c~w _e_d.-tt) r.gn?.alnm..se.vi~g~~.~ov~.d.ed ... ....t h ....t~. . . & g s . : : n o t o n t a i n . ~ ~
fli~.ys.~~.ith.b~.:this :..l.,LQ3.,~l.ch ....spa11...~n~~.~..i:~~:c:lu~.ing.edd.y current uncertaintvl.!.......::r:Ihi.s~ . . . I . ~SP-G~.
.................................................................. ..... 7 ~ ~.!incre.cs.edfol-h.
mea.sure.m.c=nt...un. ~.ertai.n.tlu'.!.. isreferred to.astheF.*-region.: If flaws are contained within the F* region. the tube shall be plugged or repaired. [2)E.F~.Cril;erion:_..II'thebottrnn.o.f the-u~p_e.rmost.hard~ tra2.siti.o.n
---- i.nt11.etubesh.c.ct...i.rs:....ab0v.e..tbe..m
... i.dn1a.n.e...oLthe tpb.c.sb.~.et butgt::J:.east2.0inch.csh.e!.owthr;.tc1.p...oft&
sec.ol?cl.a.~.. faceof'thetu besheet,.! henll!.. fla.~s!.o.cstted be1.0~~. 1,6!inchesfrol.~. the...h~ttonj ....o f t he...u~~e_~.n..ost hardroll transition (,not including eddy current uncertainty) may be allowed to remain in service provided the tube does not contain any Ilaws within this 1.67-inch span (not including eddycurrent
........................................... .... uncertainty)- ................................ T.b:1:1s.1.6.7.-.i..n.i.~_h__sxil.n (incrca.s.ed_fi>r ....rneasurern~. .nt..u.n.~.~.rt._a~t~.~...i.s...refe.
th e..E..!reglon!.......lf ..fl1.ws .... ....a . -cpnta-inedl.... ~ w.ithi:.n.!!x-EE*. reEion,the.tube. ..shallbeplugzed orrenaired, .... . (c) 'I'he lbllowing Alternate Tube Support Plate Voltaee-Based
&p.~~i~..Cri t e-rk-may..beapp_lieda~.anaIternative..t~o__the~_de~t.h b~~~:d~..~~it~l.i~_:~n..~.~.~.~..if~~~~~~~n.~~ ...For...re~~.mmssssof tln&.tub:L.&.cte.ddby...~~re.do!.n.~~~~~t.gly..g~i._a~~~y .ori:.nted ..~.ut.s.id.e diameter stress
......................................... .- .......... c ~ o s i ~ crackn . . k c o nfined wi.th.in...!h.c thickness .oftu~esu~12ort..platesthe~!u~ei..n~_or~~~air .!.imitjs as follows:
(1) If the bobbin voltage associated-with the indication i s l s s tb:z~2.rKKKKequaltn...2,0 _Vo!ts,the ..i.nd.iclati.o_ni.~..aI!c~.w.~d. . .t.o remain in_.se.n/-i.~t:~. ...
.......- If the
- 42) . . bobbin voltage associated with the indication is .
greater..than2A,... vplts,thet&e...s.h.aI! .-be...~!~~8ed.-or repaired unless the voltage is less than or equal to the upper voltage repair limit (,'calculatedaccording to the
~nethodolo.g_vin (iL95-05 as suvplemented~.anda rotating pgn~ake. ...~o.il.~Lo.r _c;o~np.arahlg.~exam~lrn.g_t~.~~n..technjqueldcr:s Prairie Island Unit 1 - Amendment No. 4-58 Units 1 and 2 5 .O-2 1 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5
!lot &tecta...fl.c??~- ...!.n..thislatter case,.. the .indicati..o~.mav
-remain in s e r v..... .h Prairie Island Unit 1 - Amendment No. 158 Units 1 and 2 5 .O-22 Unit 2 - Amendment No. 449
Programs and Manuals 5.5 Prairie Island Unit 1 - Amendment No. M8 Units 1 and 2 5.0-23 Unit 2 - Amendment No. 449
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) -Program (continued)
.(3+If
- an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits in Specifications 5 5.8 .
. 2 o M and 5.5.8.c,.2(c)(2)above@+ The mid-cycle repair limits are determined from the following equations:
Where: Prairie Island Unit 1 - Amendment No. 444 Units 1 and 2 5 .O-24 Unit 2 - Amendment No. 4-49
Programs and Manuals VURL= upper voltage repair limit VLRL= lower voltage repair limit VMuRL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMUmand time into cycle At = length of time since last scheduled inspection during which VURLand VLRLwere implemented Prairie Island Unit 1 - Amendment No. 44-8 Units 1 and 2 5.O-25 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) r . Program (continued) CL = cycle length (time between two scheduled steam generator inspections) VSL= structural limit voltage Gr = average growth rate per cycle length NDE = 95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by the NRC) Implementation of these mid-cycle repair limits should follow the same approach as described in Specifications 5.5.8.c.2(c)( 1) t&4+a"/e'/'.\and 5.5.8.c.2(~)(2)above(+. Note: The upper voltage repair limit is calculated according to the methodology in as supplemented. 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 11aws of any type (g,~,volumetricflaws, axial and circumferentd.cracks) that may..be..~re~sent.~a!.ong_t:he length oflhe t~lbc., fm.mrnth.e tu be-kt.u.bcsbe.~t
... ...we1.d... at thctubc. .h!..et..tothe.. tube-to-tybes;hgg.~w.e.@ath~~t.u bg:_outl. et,g:nG.:.tl~atmgysati,s f:vthe. a.pgJi.adb1.g.
tube re~aircri.teria:I ~ t u b e.re~.aired...~.~...s!.eev..i.n.~..~~~...~orti.
. s of..tk original tube wall between the sleeve's joints is not an area rey uiring re-inspection. The tube-to-tubesheet weld is not pal* of the tube. In addition lo ~nectingthe.~~e.~~irem~ents.~ufld~1.~..d~2~~.~dLand~
Lbs-.insp.~.~t_ion....s.c ~.~e~k_s~ec1:~t".th~1d~~.:dnd...i~s.pect~i. ngxt SI; i n s p g t i c ~ .An a s s c s s m e n t o f d e g r a ~ o gshall bc fierli~r~ncd Prairie Island Unit 1 - Amendment No. 44% Units 1 and 2 5 .O-26 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 susceptible and........................................
........................... based on........................................
this assessment,.................. to determine ............................. -. which ins_ne:tion . methodsnecclta...be .e!.n.~.!.~~ed~~~P..at.vhatl.o.ati.on_s, ...
- 1. l n s ~ e c 100% t of the tubes in each - - SG during the first refueling o~~.1.a.~_e..~o~lou~.i~ng..S(i_r.~~!~a.c~e~n.~nt~
2 ......... -.FoK.IJ.nit ...l
. .....S-Gs.s...i..nsp.ggtz,~.:QQ% ....of .th.c..tubssa t s:q~~.ntiaI...~Periiodd~.
of..l..44, @.82.-7.2.z ...... ....annd>.th_ereafter,6O ...e.ffecti.v.e .....fi!.!..,po.~e.~ ....!!?.on.th.s.! The first sequentialperiod
............................................................................................ shall be considered---to begin
-.......................................................................................................................... .......... after....................
the first inservice inspection of the SGs. In addition. inspect 50% of the tubes by the refbeling outage nearest the midpoint of the period and the reinaining_50%bv the reiue1i.n.gouta~enearest the end..d. t .h....... p.eri.0<1.~
.... ~ ........No.....SSGGGGsh.a!_l ...o.p.erat.e ...formc)reth.an 7.3 effc.ctj.v.e...h !!po.w.g~:. ... ..!.n~,nth.s~..or~..thr.c.e ....rr~fu.e.li.ng.~.ut.ag~.~
[whichever
................................................. - is less) without s being ins~ected.
............................................................... A .................................
- 3. For Unit 2 SGs, inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period
.shall be considered to begin alter the first inservi-CG .
. . inspectio.n-.of ttl.e.SCJ:s ..... Nn ....SG sha!! .....gpgrat.c-mo)r.ethan..24::l:gct1.~1.e
... ... .: ...ful! nower
*onths.or ~ ~ ~ ~ f u : ~ ~ ~ . n . g g g ~ . ~ t g a .cvi ... g ~tthhoou~::zt,t,g.i.inng
~~~y~ii~1~.e a . During each Unit 2 SG inspection (every 24 effec,tive full power months (EFI'M) or one refueling outage (whichever
~...!.~s~~.a!lt.ub~~~.withi.n....1:httS(;~hkh.ha~ve.h~
Ef:.*.._criteria arz.p!..ii~e~.ci~y.i!!.l.b~. ... .ins.p.c:t:d. in.th.c I;* and E.F* .... ...... .... ....... rggi.ons...of.ti?e..r~!! ....expanded. . . r ~ ~ iT11.e ........ ~ ~ .....r.~ai.121! n . ~ -ofth.ese tubes...be1.o.w..the...E'..*and E.E.*... rerzionsdonot-needto..be: ..... . inspected,__u_n!.ess...th-e~e,..is asleev..Sorportio!~p_t:.asleeveJ that extends below the F* or EF* region. (b). Implementation of. the " . S(;t-ube allernate repair criteria discuss-ed..i.nS.pe.clf!-~~t!.o.n-5,5,8.c,2l.~J ....re.~uir.esssaaal..O.Oo p.er~-ent..b_obbin cm!....lnspectii).n....f<r.r.
.... ..!~.otttt.l.e~AAAaar?ddddc.o!dd..!
sutluort
^ . .* -.__
date
....... ...- __ ...........intersections
" - ...-.... down to the lowest cold leg- tube U su~p.ort~Jiite ....~ . i t..k-n-o~~vn h .....outs. .i.de...diameter... stressc:orros.io*
cracking (ODSCC) indications. The determination of the lowest cold leg t&e support plate intersections having OIDSCC 7 .Y
...rndicd.ions shall _behas.e.d...o.n...tbeper~m:dnc.c
+
.... ....o-f at Prairie Island Unit 1 - Amendment No. 44%
Units 1 and 2 5.0-27 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 leasta..3.percent rando1.n.samnlingo.ftub.es....i_n_s~ected
... ver
...c!.
their
...... hll length. -. . ..
c . SG tube indications left in service as a result of application of the alternate repair criteria discussed in Specification 5.,5.,.8,c,2):I sha!! ....b.einsne~ted
. .... ...by-.bobbin . .coil p.rob.e...e.v.cr~
24 E.F.P.Mor...o.n.~...r.ef~_e.!in~~ut.~~eS
.... ... ..whih.e.verisIe.s.s.~.~..
4.. If crack indications are found in an
........................................................ SG tube2
............................................ then the.-next ......................
inspection for each SG for the defrradation mechanism that caused the crack indication shall not exceed 24 effective full power montlis or~~-~e:f'u.e!ing.~utdge.~h.i~~~~ve.r...i.s..!.e_ss~~. d_e1:initiv.e in fonnati on*:zs_umh:::g.~
................................................... :.:. f~~~~g~g.xg~m~.i~~ati~~~.n.~..~.f~~~ :. p ? !ledt~~~.g,:~d:~~gnib~_t~i,.g
~~gn~dgstruc~i~:tg.gt:i~~e~~~r::g~~.gi .nenegihhn~..e.va1.uati on ...i~~j~~Cateessssth.c!.~p crack-.like...i.nd.icati..on...isnotassociiW witl!.a crackSs.12...then...th.e ....
indicationneednot be...treatedas.a..crack. Provisions for m o n i t o r i n g p r i m a r v to secondarv LEAKAGE. f: Pro>!~.s.ro.ns
..... ......... ...f0.r...SSSG ttubtee.e.~~airrrrrn~.et.h~~ds.t
.... - .-... St.ea~n..g~.n.~r.z~ti~.r..tub:x:~.air methods.s!~.zill: pur.v.i.de.themeanstor-e.e.st~lb!.ish.. the?=.RCS .pr.e.s.sure b.o.un.da!~~.nte~~.i.t~....of ... S G....&!%~-. ~itho~t..~~n~o.v.inEt!~.et.u_b.e ....from selvice,......For tll.e~ur~osesqftLes~e ....S~.ec.i~'1.cat1. ?ns 2...tub_e.~.!..ufiei.~~1.s ....n.ot a repair. All acceptable tube repair methods are listed below.
I .- -- There are no ap~rovedSG tube..re.p:dir methods..fo_r_theUnit-.I S.G.s,
- 2. For Unit 2, the following are apgroved repair methods:
[a). Alloy 690 tungsten inert gas welded sleeves in accordance with CEN-629-1'. Kevision 03-I',"Kepair of Westinghouse Series 44 ann~.5.l.St.e:~~n~G~~.n.er'a~.orrrrTuubesssssU.siin.ri.. LC& .Tight-Sleeves?. Prairie Island Unit 1 - Amendment No. 43% Units 1 and 2 5.0-28 Unit 2 - Amendment No. 4-49
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)
- b. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:
WCAP- 14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (includes any exemption granted by NRC to ASME Code Case N-5 14).
- c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR. Steam Generator Tube Inspection Report Prairie Island Unit 1 - Amendment No. -!S&M2 4-68 Units 1 and 2 5.0-38Unit 2 - Amendment No. 44-94S S 3
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued)
- a. A report shall be submitted within 180 davs after the initial entry into MODE 4 following completion of an inspection performed in accordance with the S D ~ G ~ ~ G ~ ~ . ~ . . W J _Program ~ _ ~. .Ih.e~p-baJ.l
~B~Q~.-I SG) include:
- 2. Active degradation mechanisms found,
- 3. Nondestructive examination techniques utilized for each degradation mechanism,
- 4. Location, orientation (if linear). and measured sizes (if available) of service induced indications,
- 5. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,
- 6. Total number and percentage of tubes plugged or repaired to date, 7.
-- The results of condition monitoring, including the results of tube pulls and in-situ testing Prairie Island Unit 1 - Amendment No. -l%Kl-Q 4-68 Units 1 and 2 5 .O-3 9Unit 2 - Amendment No. 449 4-54 448
Reporting Requirements 5.6
- 8. The effective pluaaina percentaae for all pluaaina and tube repairs in wchSG,
- 9. Repair method utilized and the number of tubes repaired by each m - a j r method, and
- 10. The results of inspections
......................... --.- _ -.-.........................................perforlned under Specification
.. 5.5.8.d.3 for all tubes that have flaws below the F* or EF* distance. and were not plugged. 'T'he report shall include: a) identification of F*
and EF* tubes: a a b ) location and extent: of degradation. br,5. For implementation of the -alternate .........rejair criteria discussed in Specification
.-A .... 5
............................................................. 5 . 8 . c . .. 2 k l ~
7 , notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:
- 1b.If circumferential crack-like indications are detected at the tube support plate intersections,;
-342. If indications are identified that extend beyond the confines of the tube support plate,;g
-38. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking, Prairie Island Unit 1 - Amendment No. -l-%&+& 4-68 Units 1 and 2 5.0-40Unit 2 - Amendment No. 1494-53 4-58
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued) EM Report When a report is required by Condition C or I of L C 0 3.3.3, "Event Monitoring (EM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. Prairie Island Unit 1 - Amendment No. +5%-KLW4-68 Units 1 and 2 Unit 2 - Amendment No. 4 . 4 - m458 5 .O-4 1
RCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABLE forced flow rate, which is represented by the number of RCS loops SAFETY in service. ANALYSES (continued) Both transient and steady state analyses include the effect of flow on the departure from nucleate boiling ratio (DNBR). The transient and accident analyses for the plant have been performed assuming both RCS loops are in operation. The majority of the plant safety analyses are based on initial conditions at high core power or zero power. The accident analyses that are most important to RCP operation are the two pump coastdown, single pump locked rotor, and rod withdrawal events (Ref. I). I The plant is designed to operate with both RCS loops in operation to maintain DNBR within limits during all normal operations and anticipated transients. By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is provided between the fuel cladding and the reactor coolant. RCS Loops - MODES 1 and 2 satisfies Criterion 2 of 10 CFR 50.36(~)(2)(ii). The purpose of this L C 0 is to require an adequate forced flow rate for core heat removal. Flow is represented by the number of RCPs in operation for removal of heat by the SGs. To meet safety analysis acceptance criteria for DNB, two pumps are required at power. An OPERABLE RCS loop consists of an OPERABLE RCP in operation providing forced flow for heat transport and an OPERABLE G S- the APPLICABILITY In MODES 1 and 2, the reactor is critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the Prairie Island Unit 1 - Revision 4-72! Units 1 and 2 Unit 2 - Revision 4-72
RCS Operational LEAKAGE B 3.4.14 BASES APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses SAFETY do not address operational LEAKAGE. However, other operational ANALYSES 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 in steam discharge to the atmosphere assumes the total primarv to secondary LEAKAGE islgallon per
!ni~~.ute ...fr.~-m_ thefdulted SCretlzisass.u.m&. ..tcr..-inc_reasSccttoJ
. .....l . . p . g. ..
n~iin_ute~_ts~..t..resu!t.t.of..accdent..i.ndu.ced..co.nditi~_o~nS~lus 1.30eal!.<ms per dav from the intact SG. The L,CO reauirement to limit ~ r i m a r... y
............................ a-to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis. When the alternate repair criteria discussed in Specification
... 5.5.8.c.2Lclare iiwleinented for IJnit 2 (only), the safets!..a.na!.y.s.i.s..a.s.s.u~.n.es...th.~~.!ealtag.e...fr~~.n?....th.c...fau!t.ed S.Gz:Ll.s
..... ....!.m&d...to
.!..,42.~~~!!.0n.s....p.~r ....m.i.n.ute ....S.ba.s.e.d.. .o.n..~area~.tor...:.o.o!ant.... SY.S~-C!T!
temnerature_o.f:578.°F2.:.U'---.--r,,,,,, !AxKIAIGE . ..J ..J, Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid. The USAR (Ref. 2) analysis for SGTR assumes the plant has been operating with a 5 gpm primary to secondary leak rate for a period of time sufficient to establish radionuclide equilibrium in the secondary loop. Following the tube rupture, the initial primary to secondary LEAKAGE...zafet~a n a l ~ s ~ . s a s . u m n....is .... t l ~relatively n inconsequential when compared to the mass transfer through the ruptured tube. The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes thgl.otd1...gri l n a r ~to secondday I,,.EAK.A.G.Eis I .~~!.!~~.n...~.~r~.~.n~i.n.u.te....fro.n~...tI~.t" f'.~l!.te.d....S or isassumed to ir~cr-eas.~
... ...t.o. 1...ga!!.~n...~.e.r...~n.i.n.ut.s~ as-3..
r-esult...of .g~cde.nt..induc_cd. conditions ~ l u 150 s gallons
................................................ -. ......................................... .................. per day from the .. intact SG. When the alternate repair criteria discussed in Spec,ification 5.5.8.c.2(c) are Prairie Island Unit 1 - Revisio-4 Units 1 and 2 B 3.4.14-2 Unit 2 - Revision-
RCS Operational LEAKAGE B 3.4.14 allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this L C 0 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 BASES
- c. Identified LEAKAGE (continued) with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified leakage must be evaluated to assure that continued operation is safe.
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 L C 0 could result in continued degradation of a component or system.
- d. Primary to Secondary LEAKAGE through Any One Steam
-SG]
. . 8 .
Tljh..l.imlt
. .of..l50...gallons
. per.. d . a y . p . ~ 1r5=?! ~ ~
V S G is based on
~ t h .c t l l C I- I . . ~ ~ . --
7 L 9 7 p P ~ r ~ t .~r'E.A.K.A.G.E
.... I . o . n a a !
pe.r:rti>.nna.n.~-c ....N.E~....9.7-0~..,... crrittgrio.nnnin ~:am.2=G.en:rat~~.~.... P.roga~ Guide!.il?es !Ref, 3!,TT'h.e.Stea.m. .Gen~ratorl'ro~r~.m....o~e~:a,ti.on.a,! LEAKAGE performance criterion in NEI 97-06 slates. "'T& Prairie Island Unit 1 - IievisionAma&w&&& Units 1 and 2 B 3.4.14-4 Unit 2 - Revisiog-
RCS Operational LEAKAGE B 3.4.14 KCS o~erational
.............................. .&...................... prinla1-v
..............................- to secondarv
............................................ leakage. through
............................................... any........................
................ ..- one SG..........................................
shall be limited .......... to 150 gallons per day." . . The limit is based on operating experience with SG tube degradation mechanisins that result in tube leakage. The operational leakage rate criterion 1-n conjungtion with the i~nple~ncntittion or the Stcam BASES APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized. In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE. L C 0 3.4.1 5, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE. ACTIONS A.I Unidentified LEAKAGE in excess of the L C 0 limits must be identified or reduced to within limits within 4 hours. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent hrther deterioration of the RCPB. B.l, B.2.1, and B.2.2 Prairie Island Unit 1 - Revision- . - Units 1 and 2 B 3.4.14-5 Unit 2 - Revision-
RCS Operational LEAKAGE B 3.4.14 i.m~!.enle.ntedfor. iJnit_2lon!.y.I,.. the-saktv~ana!v~s....ass.~mesthe leakage from the faulted SG for this r e ~ a i method
.- ................................ r will be limited to 1.42 gallons per minute (based on a reactor coolant system teinperature of 578 70°F) 7
(,E uL. u 7 I The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits). The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). BASES LC0 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 hrther deterioration, resulting in higher LEAKAGE. Violation of this L C 0 could result in continued degradation of the reactor coolant pressure boundary (RCPB). LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.
Seal welds are provided at the threaded joints of all reactor vessel head penetrations (spare penetrations, full-length Control Rod Drive Mechanisms, and thermocouple columns). Although these seals are part of the RCPB as defined in 10CFRSO Section 50.2, minor leakage past the seal weld is not a fault in the RCPB or a structural integrity concern. Pressure retaining components are differentiated from leakage barriers in the ASME Boiler and Pressure Vessel Code. In all cases, the joint strength is provided by the threads of the closure joint.
- b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is Prairie Island Unit 1 - Revision-Units 1 and 2 B 3.4.14-3 Unit 2 - Kevisio-
RCS Operational LEAKAGE B 3.4.14 If unidentified LEAKAGE cannot be identified or cannot be reduced to within limits within 4 hours, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals, gaskets, and pressurizer safety valves seats is not pressure boundary LEAKAGE. The reactor must be brought to MODE 3 within 6 hours. If the LEAKAGE source cannot be identified within 54 hours, then the reactor must be placed in MODE 5 within 84 hours. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary. BASES ACTIONS (continued) The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from h l l power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and hrther deterioration is much less likely. C.1, C.2.1, and C.2.2 If RCS identified LEAKAGE, other than pressure boundary L E A K A t i E ~ ~ . r r p r i m a r ysecondarv-LEAKAGE,
.- to is not within limits, then the reactor must be placed in MODE 3 within 6 hours. In this condition, 14 hours are allowed to reduce the identified leakage to within limits. If the identified LEAKAGE is not within limits within this time, the reactor must be placed in MODE 5 within 44 hours.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions in an orderly manner without challenging plant systems. D. 1 and D.2 Prairie Island Unit 1 - Revision-Units 1 and 2 B 3.4.14-6 Unit 2 - Revisio-
RCS Operational LEAKAGE B 3.4.14 If RCS pressure boundary LEAKAGE exists or if primarv to secondary% LEAKAGE (150 gpd limit) is not within limits, the reactor must be placed in MODE 3 within 6 hours and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions in an orderly manner without challenging plant systems. BASES SURVEILLANCE SR3.4.14.1 REQ-Verifying RCS LEAKAGE to be within the L C 0 limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance. The RCS water inventory balance must be met with the reactor at steady state operating condition2 (stable temperature, power level, equilibrium xenon, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). The Surveillance is~nodifi.~d~b.y.~t.lvo_Ncrt.e.ss~ -e, r: Note:..l....~ t 3 t .is-ddtd
.. e~
dkmmg-that this SR is not required to be performed until 12 hours after establishing steady state operation. The 12 hour allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established. Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For Prairie Island Unit 1 - fievision&wwheM% !58 Units 1 and 2 B 3.4.14-7 Unit 2 - Revision-
RCS Operational LEAKAGE B 3.4.14 RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by monitoring containment atmosphere radioactivity. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in L C 0 3.4.16, "RCS Leakage Detection Instrumentation." BASES SURVEILLANCE SR 3.4.14.1 (continued) REQ-NgLg 2..,stites.that-th&S:R.is.nijt...ap.p:l&!bLg.:. to prin?aly.:=tos~:ondag I..,l:;AKAGE
.. . ....... . .... because
. ........................1,liAKAGE
.. of 150 gallons ner ....... dav cannot
. be.
measured accurately by an RCS water inventow balance. The 24 hour Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.
, , , , , , 1 , , , This SIi veriiles that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one 7 , 7 . 7 ' .
S(J,..b&f~1_nKth~.-~rima3/-
... tos:condaly-L_EAM.GE . . . .!i.m!t~n.sures tl?ttthe.o~era~onalL.EAKK~GE perfcmx~11.ce~ r ~ t . e r mthe...Stea.m
.... .n CJc.ner&t.~r. ...PIogra..n~is:~ng.~ ..not....m.et,.: c:omp.Uancglzlth
=:.: Lf.fhj.2-SR..i.s .:
L,CQ...3,4..I..:bSt,ca.m .....G..e.ncrator...Tu.b:g.In~~g~i tv '.'..sl~ou!.dbeaval:uitt..eLL
'The 1,5(>fia!.!onsper d.ay..!.imitismeasuredatroom...te~n~er.c?t.ure
- as described in Reference 4. 'T'he operational LEAKAGE rate limit Prairie Island Unit 1 - Revision-Units 1 and 2 B 3.4.14-8 Unit 2 - Revision-. 'I4 9
RCS Operational LEAKAGE B 3.4.14 applies.to I.:!.Iris.AKN X t h r ~ u l r h..an_v..!ne
.... ....SGr !.f...i ....... .t...isSs!~ot ...practical to assign the I.,EAKAGE to an individual SG. all the nrilnaw to secondary LEAKAGE should be conservatively assumed to be from one SG.
ThgSur.v:i!!&nc;.~.-i.s .....modified.b y a Not&. ... ... ..whJchstates.t!xit.:ihe Su.wci!lan.~e ....... . i s .not rcc3;uir.cd.to. - - .... .- b:.p.crfor.med..unti!. ......-...12h~uxsaftc"r - . cstab! ..i sh1.n.ent.of-:stt=ad_v ... state...operati.on.,. .E.orKCS..primarv. to secondary
.............................................. - I.,EAKAGE determination.-............................................................................................................
........................................................................................................................ steady state is defined.- as ........
stable IiCS pressure. temperature, Dower level, pressurizer and makeup tank levels, makeur, and letdown. and KCP seal iniection a.nd.retarnm...fl.o~.y_~ The.S.u.n.~i!.l.a~~.c.e....Fr.e~.u,e~~.cyY.~~ .f-...7.2....h~~~rs ....is...a_...r~~.s.o.nab!.c"..~i~?t.e.n!~! t.q ne.nd primarv... ... toseco.n.daY...I:lE.~.A_K_~..GGEEEE.an.d..reco.~~~.zes i m o .fiance-ot'earlx ..!.eakagedetect.i.o!l..in._the~~~.ve~~t.io~~...~f..tc.c:.id.e The primary to secondary LEAKAGE is determined using REFERENCES 1. AEC "General Design Criteria for Nuclear Power Plant Construction Permits," Criterion 16, issued for comment July 10, 1967, as referenced in USAR, Section 1.2.
- 2. USAR, Section 14.5.
,3,_NE:197-06, "Steam C;.e.neratorEmgra.m. C , . ~ . ~ i d . g 1 ~ ~ ~ ~ e ~ . ~
BPRI. "Pressurized Water Reactor Primary-to-Secondary Leak Guide!lnes,Y Prairie Island Unit 1 - Revision- . - Units 1 and 2 B 3.4.14-9 Unit 2 - Kevision-
SG Tube Integrity B 3.4.19 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.1Y Steam Generator (SG) 'Lube lnte~rity BASES B.ACKGRO1JND SSteamgenerato.r(SG) t u b a a r e sinall diameter. thin walled tubes thatcarry ~ri.ma.~=c.oo!.ant .... tI~ro..ugh
.. ..thg3rima.w...t.o-sgg.o.ndar~ ....
-cx:ha=es,. . TheSCrtubcs. . have ..a...numberof ..imqp-tant ... ...safet.y functions. Steain generator tubes are,an integral part of-the
........................................................................................................................................................................................................ rqactor coolant pressure boundary (RCPB) and. as such, are relied on to maintain the primary system's pressure and inventory. 'I'he Sti tubes i . s ~ o l a t e . e r ~ c t i.C:1~.~~~n.pPr~~d~~.c~.sssiinnthht'...~rir?ll.a v.c he..s.e-w.n&.systemLLL.In ..additi.~n,as_par tc~,!~..the.~RCPB.~..~
tu.be.s ar.c unigu:inthatthe.y . act-3sth.c!~g~)t:::trg.n.s.fer
.... .... ... s.u~facebet\~:.cn tl?gprim.i~ry.::gnd-..s.ccor?daw ....syste!ns~-~- .~.e.n~~~.v~~...!~.e~~t:....fr~~-m.. ....pPIimm.a,ry SYst e!n,..._TL'l?isSpecific~tion..addresses.o!1!.~. the--R.CPRint%ri.t.y function of the SG. 'I'he SG heat removal function is addressed by L C 0 3.4.4, "RCS Loops - MODES 1 and 2." L C 0 3.4.5, "RCS
.Lo__op..,.%?LCQ.3. ..MODEE 4.6. E3.~RCS~..Loop.~~~D.E....4~~.a
_ 7 -"RCS L u -x - MODE 5 Loom Filled.!." 3.4
. -2.- L-SCr.t.u.b.~ ....i.ntegrity mcangthlat.the....tu.b~sare:p.ab.l.e
.... ....nf .~Grfo.~ln_.jn.g thEir..inten.ded.-RClp.Q.~afet~ fwc!ion ....c.on.s.i.ste~~t
..... ...~v_ith...th.e....! ..i.c:.e_!!..$:..i!!.~:
basis, including. applicable regulatory requirements. S t ~ ~ ~ . g e n e r a t o r . ~ u b ~ ~ s to s u%varict.y b . i e c t d.de*gradation mechanisms. Steam generat0.r..tub.es..lnayex~eriencetube
&grad . . atk,ns.lgtgdtg=.g~j.rrc,si on..phh~:nn~~.~nngEn~~~.~s.u cha swa.sf11-gc:
pgt 1ng$. ..igt.ergran.u.j!a~ ~i.tt'a~k2~za.nd.Bgggss~~.~~rras:i~m:gyac;kingk~h with ..mechanlca!.!.v.~.nc!ued...phey?~mel?a..s.uc~ ;sdentin;and wear. 'I'hese degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are usedtomnageSGt~bede_g;rd.daG.on~ S.~.e.c.ifi..~.ati~.n .Z.,-j. 8,."Stea~n...Gzgn.er. dtor.(S.Gl...Program.." r.c.qui.g.gsthat .... a_.pro~g~n.=b.e..establis.hed a.ndin~~!c_m=.g.nted. to-ensurethat SCituhg .... integrity is maintained. Pursuant to S~ecification
..........................__ ........................................... ......................................................................................... tube 5.5.8. .......................
integrity is maintained when the SG performance criteria are met. Prairie Island - - - -- -- - -Unit
- 1 -Revision Units
-- 1 and 2 B 3.4.19-1 Unit - Revision 2
SG Tube Integrity B_3,4.,..1..9 BASES OKKGKXJND -
. ...--.-----a,.------ Th.ere.are..th:ee.SG_.~e.rf~~.m.a~~e.cr_it.e_r.i.a~
.............................. str.u_cttul:.a.!
.... ....i.nttegrity3
-.....C.. continued.1
........... ._.... accident induced leaka
. A-2 e -and .o..1.erational-.-I.FAKAGF The SG L...ZL ?.!
performance criteria are described in Specification 5.5.8. Meeting the SG performance criteria provides reasonable assurance of maintainin? tube integrity at normal and accident conditions. API'I ,lCABI .E The steain generator ...................................................................................................................... ...- -. tube ruoture...(SG'TK)
.- .... accident is the ..................................
limiting SAE'EIY A -. design basis event for SG tubes and avoiding an SG'I'R is the basis ANALYSES
-- for this Specification. The analvsis of a SG"L'Kevent assumes a boundin? priinarv to secondary LEAKAGE rate p,r_e-ater than the og..e.r.aliorna!L.EEAKA.GEErate .... ...!.i.m.itsi.n...LC_0_-34,.!4.1 '1F3.C..SSSQ~Perati.Q~.aI LE.A.KAG~.~~~!~~h.e...!.~akage...ratte ...a.s.s.o.~i.ated.-.witha-d(ru.b.1.e.-.end.g.d r u ~ t u r....e~fa-Anz!eee.tu_tr,ee,....?~!~e.accideen_t~ana!.vssiiss.s~orr.raaaa SG-TKassumes thecont~ m.. illated...sec.on~ar~....t~. .u.iPP.Pi~.S.~eL~~a~e_d...~o_theeeat.~.~.os~!!
atmospheric steam dumps. The analvses-br design basis accidents and transients other than a S.G.TR..:~.s.sum.e...the ...SG..t.u.b.e~~in.th.~.i~str.u.~t~~.ra! int.egrity...l k , . - t k . ~
.~r.e~~asSs~.rne~~~t~t.o...:u.~tur~3
-- Jn th.es.eeea_mI~~.e~.,~t!~eesteeam
.... ... ....di_s;ch~r_ge to. the atmos~her
. e_...s..b.aied-..211...thetota! ....~ r.m.a~...t.o....s.e~o_n~~.~
i. r,ElzK-A..GEoX lga!.!.on ...~.er....~.i.nute~.fro~~.~h.
... e... fau!.teddddSS or is assumed to increase to 1 gallon per minute as a result of accident induced conditions plus 150 gallons per day from the intact SG.
m e n the alternate repair criteri.aaadiscussed inSpecific-51.S..,8.,~.~.2(c_l..ar.e i.m.~~!.e.m~.nt.ed.
.... f ~..LJn r it.. 2...S.o.n!~.I.~.th.e~sa.flet~~..a~.a!~~
assume...th.e.!eakagefrc~n...t .... h.e...fctuJted. SG....!sllmlt.~d....tq .1..,42_~d!.!.o.n~.gs~ rninc&I!mM ....o ~ a r e a ~ t o ~ ~M... o os~gtenl l - a .te!nnerature .....of-518 O.F'l.: Whee..a!te~at.e .. ..criter.i.aa_d_I. .s:.c.u.s.sed....i . n . . . S ? e ~ ~. .. fc.1~2. ! c ) . are applied to axially oriented outside diameter stress corrosion crackingindications._the probabilitv that one or inore of these indications in_-.anSGwill burst under postulated main stean1lin.e break....c;i)n_clitio.ns ss.ssshhaI11.bbee...1.~.s.s...thaannnnn.1..E-.Q2..~ F'1:airle !.s!.an.d .... . . Unit 1 - -----Revision llnits 1 and 2 B 3.4.19-2 Unit 2 - Revision
SG Tube Integrity B...3..,.4.,.1..B BASES AH~'I..,ICABI For accidents that do not involve fuel damage, the vrirnaw coolant smE:' Iy F' activity level of DOSE EOUIVALEN'I' 1-131 is assumed to be eaual a ANNALYSEs to or greater than the L C 0 3.4.17, "KCS Specific Activity," limits. ccont.i.nued) For accidents that assume fuel damage. the ~rirnarvcoolant activily is a fun.~ti.on-of the amount of activitv released from the damaged fu-G!.,. .. The...d~~se..conse.qu.e.nces. ~ftl?ese.e\ic.nts-are-.wi_t.hi~~.=the....li_!nits.. GDC...1..9(.Ref*...2JA..l..Q
... ... . .... ....C.F.R:A.~Q.0.2.L.~g~~ ~J..~J~~~=C..N.R:C~.~~~
!icengin~.b.asis.!e!g,.:. a.sma!!fractioll.of'these!imits!.
Steam generator tube inte-gritysatislks Criterion 2 of 10 CFK 5-026@)(2 )(i&.
. . .::.=:.:::::::.::.:::.::::.= Thc-.l,C.Org:g.ui.r.c.s...th~.t:..
.... $.G tub.e..integritv....b.e.._n_?.aintainncdgk
... .=::. Th.c...L..C..Q a..........1so req.- ui .......................
- res_th .. at a1 1 S G @ bes thaJ-~._a_tis.fy~~t~~~...re~di~r~crite.ri..a.. be plugged or repaired in accordance , .-- .... with the Steam Generator Program.!.
During an SG inspection, anv inspected tube that satisfies the Steam Generator Program repakcriteria is repaired or removed from se.wi.ce..by~!.ugg.i.ng
. . If a..tube...was..~d.et.er~ni.n.e.d...t~~...~t~&th~...re crrtena.butwa.saotnlu~gd=:~rrgpa.ircd~.the_ ... tube.ma~ ....still . . t u b e integrity, In the context of this Specification, an SG tube is defined as the entire length of the tube, including the tube wall and any repairs n.3.d-etoif,_he.t\.~~e.tube-to-t_ubesh.e.e~ weldatthetubeinletand th.ctube-to-t.u.b.e.sh
......... ... .e_e.t...~~.e!d-.at..tl?:.._tu.h:
-. eutl.gt2
.... .::. The .t~h:~t.o-t.~.besh~et
=
xe!.d is n.ot.co.n;idercd...part..ofthe..,tlrbe:n o r .is_t,!l-~
.... ... ....r.cgi.o).nof3ub.c....
be! .OJ~...the ....E.* ....al7d.~.~F*...~_e~~.o0n....Ce~.ce~.t...is ....~.oted ...be,o~1.~...~o.1:..tt_le portion of................. the....tube between sleeve ioints. When an F* or EF* region is repaired by sleevin-q. the entire sleeve is considered part of the tube. Prairie Island -- - - Unit 1- Revision Units 1 and 2
- - B 3.4.19-3 Unit 2 - Revision
An SG tube has tube integrity when it satisfies the SG perfor_mance criteria. 'The SG performance criteria are defined in Specification 5.5.8,"Steam Generator I'rogram," and describe acceptable SG tube performance. The Steam Generator Program also provides the LC0 -evaluation process for de&rm&i_ng conformance with the SG (continued) - pxfannance criteria,
'There are three SG -
............................... performance
-.................................................. --criteria: structural integ~-ity~accident induced leakage. and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.
a~ticipatedtral~sientsincluded in the design.s~cificati0n.Il'ubc burst is deiined as, "?'he gross structural failure of the tube wall.
'I'he condition typically corresponds to an unstable opening dis&c~~n.en~,.I~..~g.~. >.. o.pel?ing.. arcxi increa;.edinrespon,s..eett~ .. ~onsta~lt p.r.es.s~lreI_ac_c;olnp~niedby.d~~tii~Dlastic) teafinx-~f.
... thetube 1~~~tc.ri.a.1 ....at .th.c.ndso>f.tb.g...d~grad.atipn Tube_.. ce!!apse i S dee.ned as,.~E.~r...th.e...!.oad....dis~.!~cerne~7t.c~ . o .?
given structure. collapse occurs at the top of the load versus disulacement curve where the slope of the curve becomes zero. The s&u~tua!-...inlegritv.z_erfma.ncecriterion gr_ovidmui.dance~m ass.essi.ngloadsthd..have a s i ~ n i l k a n m.on-b~l_rs~ .. t c o l l a se,In ~ that..cont.e.xL:- ..tha...t.grm...;bsig!!ifi.~llc.: -- :...isd.g.flngd....as...t~A.n.~(i.c.c.id~.n~ Ioading..~~n-ditien . 0 t h ~r...th3.n . d.iffer.~l~tia!r.c.s.s.u.r.gis..~onsid.cred
.... ...I?. :
significcmt
....... 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 cslab3ishe__ddd7 F_c,rtrbe. .... . i n
.~xunl:feren!d....d.cgradati~n.,.axhUhe.r~_na.! ... ..load.sare.cldssi17ed as scctznd ary doads., .T;~~~cil.cumf~s~~e.ntig.1..:: . ddegradgt.io.nthg.zcck4%d.ii=~Eadi~).n of axial thermal 1oads as prin1ary_~?.rse:c11~9Iry
.-. .. . .-.- .!oa.ds.willb:
evaluated on a case-by-case basis. 'The division.between p r i r n a ~ Prairie Island - Unit 1 -- Revision
-- 1 a ~ 2 ilnits
- d B 3.4.19-4 Unit 2 - Revision
SG Tube
-- - - Integrity B 3.4.19 a ~ ....idecz darv.clasSlfications...~~i.!.! ...be-based....on.detai!.ed-_an.a,!.~ s.i..s andlor testing, ...................
1,C:'O Structural integrity requires that the primary membrane stress ( conti.nu~.dl ...................... iinter.l.ss~~y.~.nI?a~ubb~ .-n~rttteicce.eed.th~~ie.l d. sir.ength....f'~~r. ..a1lASMEC_o&, .... Sectio11.11.1,.S.cwiicceLeve1A l . n a ! o a t ..cs~.n.di.tion&i~l.d.
$g.pi.i:..I,ave!==~. ~.u.P.R.G~...~~ abnor.n~gl~cgt~diti.gns
. . 1 transient~included in.t11.~d.c.s.ign~..s.p.c.~ificat.i.o.n~~
.... :.: Th~sr..n~l~~~.:gssafeQ..~d.~to.r.s ....and applicable
.....* *........................ design basis loads based on ASME
................................................................ Code. Section 111, Subsection NB (Ref. 4) and Draft Regulatoiy Guide 1.121 (Ref. 5).
T~~c~.accid~nt..hducedleakage_~erfomanc.e.cr~ite.rion.~e__nsures.t;hat primary .. to secondary LEAKAGE caused by a d e s , n b a s i s accident, oth-e.rthan m$.G.TR
.ThgIilccid:.nt....analysis...ass~l.m:es
.................. . is.witl?l.n..the.agcid.ent thata.~c.i .d.e t ~...indu
..a!u.!ysi.~as.su.n~.~ti.t~ll.s..+.
t cedl.. kak3g.c...d:.o.es not
............... exceed those discussed ... in the AF'PT..,ICAHI,E
.................................................................... SAF1?17Y ...............................
ANALYSES section above. 'T'he accident induced leakage rate indudes any primarv to secondarv LEAKAGE existing prior to the accident in addition to y r.ilnary~ec:Ond.cl:ry ....L.EAKKAAGIEEEEEinduceSt during the .... -- - accident.
~ h ~ . . ~ p ~ ~ . g ~ . c ~ ~ ~ r ~ = ~ I I.ccr1t.er~:onnpr~)yidg.s ~ r ; , ~ . K ~ ~ Gan .... EEEEpP~.rf~
observabielndiati onof S G tube .condi-tiol?sduring ..~.!.c?nto~erL?tio~~. I'he limit on operational LEAKAGE is contained in L C 0 3.4.14, "RCS Operational LEAKAGE." and limits primary to secondary LEAKAGE . through ... -- any -- .....one SC; to ..1 .. ......5
.......0
... g dllonsp ~ ....d_n_y_. r -T.h.is..l.imit i s based .o_nt-he-as~u~n~tion...t hat...asb~gle..srac.k. l.eakixt_t_.s ....a mo.u.r!t
\su.l.d. . . ~ ~ ~ t p r ~ > pan.SCi.T.R...:u.n.~.er .... a ~ t ~ . . . . t ~ ~th: stress ~ot~&!~~.nsof . . ... a
.,O..C.A...or..a ...m.air7 ste.am 1.in.ebre.ak......lfthisamount
.... ., ... .. of LE.AKAGEis .
due to...!.n.o~~-.t~!~n...one..crak.,..the ....crack . . . . verysma!.!.,..al1d the..above assuinption is conservative. ~~ILICABlLI'l'Y Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Larpe differential pressures across SG tubes can only bg._e_xeriencedin...M0.DE.L,2 3.,.~~4.. ., &ai-ri Island Unit 1 -- Revision LJnits 1 and 2
- -- -- .- B 3.4.19-5 , iU
-- n
SG Tube
- Integrity ................
I$..3..?.4.,b9 K C S c o n _ d i t i o n s a r e ~ e e e e . i i n n n n5.2nd.. n n M6..th.an MO~~!L~Ss durin~..R/lODl-)S
... ....................................... 1. 2, 3. and _ 4. . In_ MOIIES 5 and 6. *primary to
&'I'LICABlLi'I'Y secondary differential pressure is low. resulting in lower stresses and (continued)
- rcduced potential for LEAKAGE.
ACTIONS ................................................ TheA.CT!0N~_re~modi~fi.cd....b~~.a._N~arili.ing.lb~~th~o. may be.cntgr.gdindcaeudcntly....fore.a..c:.~.h:SsG...ttub..~ .... .,..Th.is..i:g.~gcc;~.~t~~?!.c because the Required
-. .- -........-Actions provide ap~rgeri.git.e
..................................................................................... . c.o.mp.c.n.sgtorv acti.ons..for..e~ch...affected...S.~...t.~be, Com_ml~ :Iis....withth.eeeefi:cluired Actionsma_va!,~\?iforcor?t.in~~e.d... operatiwon,a!ld subsecluent...affect e6!
SG tubes are governed by subsequent Condition entrv and application of associated Required Actions. condition. A a ~ p i j e s i.it..k . f discovered t!~.at . ..one..01 moreSG tubes ex-aminedinaninservice .ins~ectio.n..-satisf~...thetube~e~air criteria but were not plugged or repaired in accordance with the Steam Generator E'ropram as required by SIX 3.4.19.2. An evaluation of S (i....tub.eintcg~.i.t~~.o!~~he .... ._a~ft".cl:.e.ddddt~~bgI_s~. must-&.. m ati.e..Stc:a.m g~.~~~~t.o~,t~~hg~~int.g.grit~;,:i:s .: b.asedgnm::ti.n.g t ; h ~S.Glpx.f~~.~~gl~.cc crite-isdess.r.i:b~&&l?: . S t . ~ ~ . m ~ ~ ~ ~ ~ ~ ~ g n e r The... a t ~ S..G ~ . r.~epalz ...P.~~~r:.~~ criteyiadlefi.ns...!in!i.ts ...on...S.
.......... Citubs..degr.gdat.i.0.n. ..thatallow....for f l ~ ~
ero~~t!~...b.e5~esn jEspect.~.ons
..... ..~?ihi.!.e....st.i.!!~~.o~viciln~.astsura~~.c:.e. ...t~~c?t....the SG perfortnance criteria will continue to be met. In order to determine if an SG tube that should have been plug-ged or repaired babeintexri~ii *.. anevaluat:ionm__ustbe.sompl
, . et.cdthat sl.e~n.cr!?.strat.e.s. ... that.-- the....SG.p.e:forma.nc.~ ....crlttcrl.aaajy..i!I. cor.nti!!:gs~~bg mct
- until..t l next.... . ~ refu.cling..s.~ita~eor ...S.G tu.b..e...i.n~p~~tio.n
... .:.........T ~ . ~ . . t , ~ b e .
integrity determination
........................................... _..._._ ... is based on the estimated condition of the tube...at-t!>etil.n.e. Lhe ~.ituario.n ... ...ni.s ....d1.s.c.o.vered andth.e.esti!.n.at~d ...gr.owth of the degradation prior to the next SG tube inspection. If it is determined that tube-ht.egritv.is. notbeing;.-maintained,.....C ~1nAtionB ap.plie&
)'rairkisland . -- - Unit 1 - Revision LJnits -- -- 1 and 2 -- --- B 3.4.19-6
.- - Unit 2 - fievision
A Completion Tiine of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with an SG tube that may not have tube integrity. AC'I'IONS -- A. 1 and A.2 (continued) T-!~th_e.cv_'cllu.ationdetermines_tb_atttthe_affected tube~sl-huxeL!ubs integr_it.y.,._Res.ui:ed..A.cti.oo!! ...A,,2....al.1~~s~.>!.ant~.c~.~~~~ti~~~n...t.oc. unti l..the...n
.- e.xtrcfuding auti~g.~~r-..S.CJ
... .....ir~.~p~gcti..o.~.-.p-r-~.vi.d.cd.t~~g inspection
............... . -- interval continues to be supported
- ............................................. . by an operational assessment that reflects the affected tubes. However. the affected tube(s) must be plugged or repaired prior to entering MODE 4
.fo.llowing the next refueling;outag~~~1:.S~iins~e..c.t.i.o.n.._~Th1~
C..~n..!.e.in.... T i . m . . . c c . b ! .. i ne . . . . . n . . . . i !.the . ne-xl b:y::.the .~)pPPer~_fjoggl~ss:g.ssmg~!t,. lf the Keuuired Actions and associated Completion 'Times of Condition . A are not met or if;_O.t-ube integrity is n o t k i n g m.aintain__e_d,_thereact~x.muAMri~~~htt,MODE..3-wit~ia.C>l?o a.nd...M.QD.E...S. .~~ith.i1?36lncrurs,
'l'he allowed Con~pletion'l'imes are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant sy ste~ns.
SIJRVElLLANCE SR _ 3,4.19.1 During-shutdown periods the SGs jme_inqcctcd as reyui~cc! by this SK and the Steam Generator Program. NIII 97-06, Steam Generator F'rograrn Guidelines (Ref. l)? and its referenced 1:l'RI Guidelines. establish the content of the Steam Generator I'rogram. Use of the Steam-(jenerator P r ~ a a mensures t h a t ~ h ei-nspection is appropr-iate and consis~entwit11 acsgpted industw practices. Prairie lsland Unit 1 - Revision LJ~lits1 and
- --2 B 3.4.19-7 Unit 2 - Revision
SG Tube Integrity B 3.4.19 BASES s!JRV]:ZT!ii!d./2NCE . ... ....S.R .::: 2.42!19z.1 : l
. [.g:o:ntiuu.~dj.
Kl ~(?IJIKf..~MI;~N'~S; During SC; inspections a condition monitoring; assessment of t h 5 . m tubes is performed. 'I'he condition monitoring assessment determines the.... "as hund" condition of the SG tubes. -- The_pu.rp.os.e. ...of th.econdi~i>nm~>.nito.r.Lnn aszessnlerg-i.stomsura . tb.attheXi ... pga-f~)ma.n~.$ :::. ggit&&.i:g .: h:av.c...bc.cu n~.xtf<j.b.:th.c=:pr.cvi
..... .:.: o.u.~?o.pa-ati.nx pgria->g<.
7'he Steanl Generator I'rograin determines the scope of the inspection and the methods used to determine whether the tubes contain - .-..-flaws
.-. -sat-isfyhgLh.e tube repaircriteria l . . . i i i i l n ~ ~ c t i _ o . n . ~ ~ ~ ~ P e Ci.,e ,. which .tu.besoram_l'tub-i-nxwithintheSCi-.axcto-be . . .
inspect.ed)is.3 Lujmctinn nl-.cx~.st~..ng;a~~otential..d:ggradation
-1.0.catio11.s
...-- .p..-.--...----p-p-. ,.T he ~t~~!n~.~~.~.nerat~r...P.r.ogra.m
... ....al..s.oO.speci~~.e.s the irlsnection
.......... .- ........ .-....methods to be used ...........
to.....find potential . ..degradation. Inspection methods are a function of degradation morpholog~~ non-destructive examination (NDE) technique capabilities. and i.nsp:.~tio~!_c~.cation.s~~ The-Stcam -- Gcngrator Program -- defines the Frcqu_ei~cy_of SR 3.4.19.1. The Frcqucncy -- - is dctcrininqd by the opcratic-mgl assessnlerlt and ither lirnits in the examination guidelines (Ref 6). The Steam Generator Program uses information on existing degradations and growlh rates lo detcrlnine an=inspection1;rcquency that pro\ ides 5.5.8 containsprescr-iptivc .- rgquirenicnts concerning inspection ingr!lals to provid~~added
-- assu-rancc that the S$ p e r f ( ~ n ~ i n c c ~ r & i a will be met between scheduled inspections.
Prairie-- Island -_ .. -- -Unit - 1- -- Revision Units 1 and - 2 . - _ - _ - - - p - B 3.4.19-8 - - Unit 2 - Revision
SGjl'ube Integrity B 3.4.19 Iluring, an SG inspection.
................................................... any insnected a .-. ......................... tube that satisfies the Steam Generator Program repair criteria is repaired or removed from 1
Specification
....... -. 5.5.8 are intended to ensure . -- that tubes accepted for co!jti.nued senii.~:~ ... .... sattiis.@ . . allowancg .
.f:lrre.rrcrrinthe flaw size...m~~.as.u.r.e.m.e!1t
- ..... .... ....a.nd...f~r..futu.r.c....flaw .. gr.oytJ.
I! . addition, the .tut>e..re~air ... .. crite13~i.ncon1~1!1.ct1.0n...~~ith .... ether
.elements of the Steam Generator Program. ensure that the SG performance criteria will continue to be met until the next inspection of the subiect tube(s). Reference 1 provides guidance for performing operational -assessments to verify that the tubes remainin~inservice wi.!!continu . ..
... -- .~.to...mc.~.t....th.~
....................... SGperformans.~
......... .... .--- ..... - . cx.iteria, s~ea-mgenerct.ttor...t~b.e...~e~r!i~.s~.~.are..
.... E!Y ..perforlnedu s i n g a ~ ~ r o v e d re_nainnetl?qdsas.descl:ibedin the.Stean~CJt=neratorPrograin.
'rhe Frequency of prior to entering MODE 4 following an SG insection - ensures that the Si~rveillancehas been completed and all tubes . . n3et.ing.tl%:g.::r.cna.ir .c.rlter~a..ar.gplugcd. or...rg.giredlprj.i2r.::to sub~~~t~.ng..th~.~~S~~.tub~.s..t~.~~gni.f!c~~
differenti a!. Prairie Island Unit 1 - Revision I Jnits 1 and 2 - - B 3.4.19-9 Unit 2 - Revision
SG Tube Intcgrity B 3.4.19 BASES (continued] REF17RENCES
-- _ 1. NEI
-- 97-06hbSte;imCJqncratgrProgram Guidcli~~cs."
- 2. 10 CFlt 50 Appendix A, GDC 1 2
- - 4. A@IE Boiler and P r e s s u r g V e ~ s c l C ~ ~ e ~ SIIISubsectic~t~
~cti~~n Nn .
- - ---- .- 5. Draft Regulatory Guide 1.121. "Basis for I'lugqing: Degraded Steam Generator 'I'ubes," August 1976, Prairie Island -- - - -- - - - -- - - -- --- - -- Unit 1 ---Revision (Units 1 and 3. I3 3.4.19-10 -- Unit 2 - Revision
ENCLOSURE 3 Proposed Technical Specification Pages (revised) Prairie Island Nuclear Generating Plant Units 1 and 2 Technical Specification Pages 20 pages follow
Definitions 1.1 1.1 Definitions (continued) E -AVERAGE E shall be the average (weighted in proportion to the concentration DISIN'IEGRATION of each radionuclide in the reactor coolant at the time of sampling) ENERGY of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant. LEAKAGE LEAKAGE from the Reactor Coolant System (RCS) shall be:
- a. Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
- 3. RCS LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
- b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) 1 through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
Prairie Island Unit 1 - Amendment No. 43-8 Units 1 and 2 1.1-3 Unit 2 - Amendment No. 4-49
RCS Operational LEAKAGE 3.4.14 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. RCS identified C. 1 Be in MODE 3. 6 hours LEAKAGE not within limit for reasons other AND than pressure boundary LEAKAGE or primary to C.2.1 Reduce LEAKAGE to 14 hours secondary LEAKAGE. within limits. 44 hours D. Pressure boundary D.l BeinMODE3. 6 hours LEAKAGE exists. AND D.2 BeinMODE5. 3 6 hours Primary to secondary LEAKAGE not within limit. Prairie Island Unit 1 - Amendment No. 44% Units 1 and 2 3.4.14-2 Unit 2 - Amendment No. 4-49
RCS Operational LEAKAGE 3.4.14 SURVEILLANCE FEOUIREMENTS SURVEILLANCE I FREQUENCY SR 3.4.14.1 .......................... NOTES ..........................
- 1. Not required to be performed until 12 hours after establishment of steady state operation.
- 2. Not applicable to primary to secondary LEAKAGE.
Verify RCS operational LEAKAGE within limits 24 hours I by performance of RCS water inventory balance. SR 3.4.14.2 .......................... NOTE ---------------------------- Not required to be performed until 12 hours after establishment of steady state operation. Verify primary to secondary LEAKAGE is 72 hours
< 150 gallons per day through any one SG.
Prairie Island Unit 1 - Amendment No. 4% Units 1 and 2 3.4.14-3 Unit 2 - Amendment No. 4 4
SG Tube Integrity 3.4.19 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.19 Steam Generator (SG) Tube Integrity LC0 3.4.19 SG tube integrity shall be maintained. All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program APPLICABILIIY MODES 1, 2, 3, and 4. ACTIONS .................................................. NOTE .................................................. Separate Condition entry is allowed for each SG tube. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A. 1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next or repaired in accordance reheling outage or SG with the Steam inspection. Generator Program. AND A.2 Plug or repair the affected Prior to entering tube(s) in accordance with MODE 4 the Steam Generator following the Program. next reheling outage or SG tube inspection Prairie Island Unit 1 - Amendment No. Units 1 and 2 3.4.19-1 Unit 2 - Amendment No.
SG Tube Integrity 3.4.19 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.l BeinMODE3. 6 hours associated Completion Time of Condition A not AND met. B.2 Be in MODE 5. 36 hours SG tube integrity not maintained. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.19.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program SR 3.4.19.2 Veri@ that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged or repaired in MODE 4 accordance with the Steam Generator Program. following an SG tube inspection Prairie Island Unit 1 - Amendment No. Units 1 and 2 Unit 2 - Amendment No.
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
- 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 or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired 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 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, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.8.c.2(c), 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.
Prairie Island Unit 1 - Amendment No. I48 Units 1 and 2 5.0-13 Unit 2 - Amendment No. 149
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) 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. For Unit 2, when alternate repair criteria discussed in Specification 5.5.8.c.2(c) are applied to axially oriented outside diameter stress corrosion cracking indications at the tube support plate locations, the probability that one or more of these indications in an SG will burst under postulated main steam line break conditions shall be less than 1E-02.
- 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. For Unit 1, leakage is not to exceed 1 gpm per SG. For Unit 2, leakage from all sources, excluding the leakage attributed to the degradation associated with implementation of the voltage-based repair criteria discussed in Specification 5.5.8.c.2(c), is not to exceed 1 gpm per SG.
- 3. The operational LEAKAGE performance criterion is specified in L C 0 3.4.14, "RCS Operational LEAKAGE."
- c. Provisions for SG tube repair criteria:
- 1. Unit 1 steam generator 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.
2 Unit 2 steam generator tubes found by inservice inspection to contain flaws shall be dispositioned as follows: Unit 1 - Amendment No. 438 Prairie Island Units 1 and 2 5.0-14 Unit 2 - Amendment No. 44-9 I
Programs and Manuals 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) (a) Depth Based Criteria: (1) Tubes found by inservice inspection containing a flaw in a non-sleeved region with a depth equal to or exceeding 50% of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate tube repair criteria discussed in Specification 5.5.8.c.2(b) or in Specification 5.5.8.c.2(c). If significant general tube thinning occurs, this criterion is reduced to 40% wall penetration. (2) Tubes found by inservice inspection containing a flaw in the pressure boundary region of any sleeve exceeding 25% of the nominal sleeve wall thickness shall be plugged. (3) Tubes with a flaw in a sleeve to tube joint that occurs in the original tube wall of the joint shall be plugged. (b) The following F* or EF* Alternate Repair Criteria may be applied to the hot-leg of the tubesheet as an alternative to the depth based criteria in Specification 5.5.8.c.2(a)(l): (1) F* Criterion: If the bottom of the uppermost hardroll transition in the tubesheet is below the midplane of the tubesheet, then all flaws located below 1.07 inches from the bottom of this uppermost hardroll transition (not including eddy current uncertainty) may be allowed to remain in service provided the tube does not contain any flaws within this 1.07-inch span (not including eddy current uncertainty). This 1.07-inch span (increased for measurement uncertainty) is referred to as the F* region. If flaws are contained within the F* region, the tube shall be plugged or repaired. Prairie Island Unit 1 - Amendment No. 43-8 Units 1 and 2 5.0-15 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) (2) EF* Criterion: If the bottom of the uppermost hardroll transition in the tubesheet is above the midplane of the tubesheet but at least 2.0 inches below the top of the secondary face of the tubesheet, then all flaws located below 1.67 inches from the bottom of the uppermost hardroll transition (not including eddy current uncertainty) may be allowed to remain in service provided the tube does not contain any flaws within this 1.67-inch span (not including eddy current uncertainty). This 1.67-inch span (increased for measurement uncertainty) is referred to as the EF* region. If flaws are contained within the EF* region, the tube shall be plugged or repaired. (c) The following Alternate Tube Support Plate Voltage-Based Repair Criteria may be applied as an alternative to the depth based criteria in Specification 5.5.8.c.2(a)(l): For regions of the tube affected by predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of tube support plates the plugging or repair limit is as follows: (1) If the bobbin voltage associated with the indication is less than or equal to 2.0 Volts, the indication is allowed to remain in service. (2) If the bobbin voltage associated with the indication is greater than 2.0 Volts, the tube shall be plugged or repaired unless the voltage is less than or equal to the upper voltage repair limit (calculated according to the methodology in GL 95-05 as supplemented) and a rotating pancake coil (or comparable examination technique) does not detect a flaw. In this latter case, the indication may remain in service. Prairie Island Unit 1 - Amendment No. 448 Units 1 and 2 5.0-16 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) I 3 If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits in Specifications 5.5.8.c.2(c)(l) and 5.5.8.c.2(~)(2)above. The mid-cycle repair limits are determined from the following equations: Where: VURL= upper voltage repair limit VLRL= lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLm= mid-cycle lower voltage repair limit based on VMUm and time into cycle At = length of time since last scheduled inspection during which VURLand VLRLwere implemented Prairie Island Unit 1 - Amendment No. 44-8 Units 1 and 2 5.0-17 Unit 2 - Amendment No. 449
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program- (continued) I CL = cycle length (time between two scheduled steam generator inspections) VSL= structural limit voltage Gr = average growth rate per cycle length NDE = 95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by the NRC) Implementation of these mid-cycle repair limits should follow the same approach as described in Specifications 5.5.8.c.2(c)(l) and 5.5.8.c.2(~)(2)above. Note: The upper voltage repair limit is calculated according to the methodology in GL 95-05 as supplemented. I
- 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. In tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d. 1, d.2, d.3 and d.4 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the Unit 1 - Amendment No. 43-8 Prairie Island Units 1 and 2 5.0-18 Unit 2 - Amendment No. 4-49 I
Programs and Manuals 5.5 5.5 Programs and Manuals Steam Generator (SG) Program (continued) 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. For Unit 1 SGs, inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.
- 3. For Unit 2 SGs, inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs.
No SG shall operate more than 24 effective full power months or one refueling outage (whichever is less) without being inspected. (a) During each Unit 2 SG inspection (every 24 effective full power months (EFPM) or one refueling outage (whichever is less)), all tubes within that SG which have had the F* or EF* criteria applied will be inspected in the F* and EF* regions of the roll expanded region. The region of these tubes below the F* and EF* regions do not need to be inspected, unless there is a sleeve (or portion of a sleeve) that extends below the F* or EF* region. Prairie Island Unit 1 - Amendment No. 458 Units 1 and 2 5.0-19 Unit 2 - Amendment No. 4 4 9
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) (b) Implementation of the SG tube alternate repair criteria discussed in Specification 5.5.8.c.2(c) requires a 100 percent bobbin coil inspection for hot leg and cold leg tube support plate intersections down to the lowest cold leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length. (c) SG tube indications left in service as a result of application of the alternate repair criteria discussed in Specification 5.5.8.c.2(c) shall be inspected by bobbin coil probe every 24 EFPM or one refueling outage (whichever is less).
- 4. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
- e. Provisions for monitoring operational primary to secondary LEAKAGE.
- f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.
- 1. There are no approved SG tube repair methods for the Unit 1 SGs.
Prairie Island Unit 1 - Amendment No. -I48 Units 1 and 2 5.0-20 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 5.5 Programs and Manuals Steam Generator (SG) Program (continued)
- 2. For Unit 2, the following are approved repair methods:
(a) Alloy 690 tungsten inert gas welded sleeves in accordance with CEN-629-P, Revision 03-P,"Repair of Westinghouse Series 44 and 5 1 Steam Generator Tubes Using Leak Tight Sleeves". (b) Hardroll expanding non-sleeved portions of tubes in the tubesheet in order to apply the F* and EF* criteria. Prairie Island Unit 1 - Amendment No. 44-8 Units 1 and 2 5 .O-2 1 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) This page retained for page numbering Prairie Island Unit 1 - Amendment No. 44% Units 1 and 2 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 Programs and Manuals (continued) This page retained for page numbering Prairie Island Unit 1 - Amendment No. 44% Units 1 and 2 5.0-30 Unit 2 - Amendment No. 4-49
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) This page retained for page numbering Prairie Island Units 1 and 2 Unit 1 - Amendment No. 4% Unit 2 - Amendment No. 4 4 I
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)
- b. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:
WCAP- 14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (includes any exemption granted by NRC to ASME Code Case N-5 14).
- c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR. Steam Generator Tube Inspection Report
- a. 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 5.5.8, Steam Generator (SG)
Program. The report shall include:
- 1. The scope of inspections performed on each SG,
- 2. Active degradation mechanisms found,
- 3. Nondestructive examination techniques utilized for each degradation mechanism,
- 4. Location, orientation (if linear), and measured sizes (if available) of service induced indications, Prairie Island Unit 1 - Amendment No. 449 4-68 Units 1 and 2 5 .O-38 Unit 2 - Amendment No. 4% 44%
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued)
- 5. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,
- 6. Total number and percentage of tubes plugged or repaired to date,
- 7. The results of condition monitoring, including the results of tube pulls and in-situ testing,
- 8. The effective plugging percentage for all plugging and tube repairs in each SG,
- 9. Repair method utilized and the number of tubes repaired by each repair method, and
- 10. The results of inspections performed under Specification 5.5.8.d.3 for all tubes that have flaws below the F* or EF* distance, and were not plugged. The report shall include: a) identification of F*
and EF* tubes; and b) location and extent of degradation.
- b. For implementation of the alternate repair criteria discussed in Specification 5.5.8.c.2(c), notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:
- 1. If circumferential crack-like indications are detected at the tube support plate intersections,
- 2. If indications are identified that extend beyond the confines of the tube support plate, or
- 3. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
Prairie Island Unit 1 - Amendment No. 4-62 4-68 Units 1 and 2 5.0-39 Unit 2 - Amendment No. 4-33 45-8
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) EM Report When a report is required by Condition C or I of L C 0 3.3.3, "Event Monitoring (EM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. Prairie Island Unit 1 - Amendment No. 4-634-68 Units 1 and 2 5.0-40 Unit 2 - Amendment No. 4.44 4423}}