ML12220A012
| ML12220A012 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/31/2012 |
| From: | Price J Dominion, Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 12-485, TSTF-510 | |
| Download: ML12220A012 (23) | |
Text
Dominion Nuclear Connecticut, Inc.
0 5G00 Dominion Boulevard, Glen Allen, VA 23060 2i) Dominion Web Address: www.dom.com July 31, 2012 U. S. Nuclear Regulatory Commission Serial No.12-485 Attention: Document Control Desk NSSL/MLC RO Washington, DC 20555 Docket No.
50-423 License No.
NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 3 PROPOSED TECHNICAL SPECIFICATIONS TO ADOPT TSTF-510, "REVISION TO STEAM GENERATOR PROGRAM INSPECTION FREQUENCIES AND TUBE SAMPLE SELECTION," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) requests an amendment, in the form of changes to the Technical Specifications (TS) for Facility Operating License NPF-49 for Millstone Power Station Unit 3 (MPS3). The proposed amendment would modify TS requirements regarding steam generator tube inspections and reporting as described in Technical Specifications Task Force (TSTF) Traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
The availability of this Traveler was announced in the Federal Register on October 27, 2011 (72 FR 66763) as part of the consolidated line item improvement process (CLIIP).
Because MPS3 has not adopted Standard Technical Specifications (STS), DNC is proposing minor variations and/or deviations from the TS changes described in TSTF-510, Revision 2, to provide consistent terminology and format with the MPS3 TSs. The minor variations and/or deviations from the specific wording/format provided in TSTF-510, Revision 2, are considered administrative and do not change the meaning, intent, or applicability of the CLIIP. provides a description and assessment of the proposed changes including: the requested plant-specific licensing basis that demonstrates compliance with the 10 CFR 50, Appendix A General Design Criteria referenced in the Traveler, as well as the plant-specific administrative variations from the TS changes described in TSTF-510, Revision 2. Attachments 2 and 3 contain the marked-up TS and TS Bases pages, respectively. The marked-up TS Bases pages are provided for information only.
The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program when this amendment is approved.
Serial No.12-485 Docket No. 50-423 MPS3 TSTF-510 - SG Program Page 2 of 3 The proposed amendment does not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92. The Facility Safety Review Committee has reviewed and concurred with the determinations herein.
Issuance of this amendment is requested no later than July 31, 2013, with the amendment to be implemented within 60 days of issuance.
In accordance with 10 CFR 50.91(b), a copy of this license amendment request is being provided to the State of Connecticut.
If you have any questions or require additional information, please contact Ms. Wanda Craft at (804) 273-4687.
Very truly yours, J. f Prdi c e,,
Vice President - Nuclear Engineering j
VICKI L. NULL Notary Pufbl Comm"wAlf of Virginia COMMONWEALTH OF VIRGINIA
)
j 140542 M
My Ccn,mn May 31.2014 COUNTY OF HENRICO
)
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. Alan Price, who is Vice President -
Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this day of\\
,'2012.
My Commission Expires: /,k/)
.1 Iz
/
Notary Public Commitments made in this letter: None
Serial No.12-485 Docket No. 50-423 MPS3 TSTF-510 - SG Program Page 3 of 3 Attachments:
- 1. Description and Assessment
- 2. Marked-Up Technical Specification Changes
- 3. Marked-Up Technical Specification Bases Changes (For Information Only) cc:
U. S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd., Suite 100 King of Prussia, PA 19406-2713 Mr. J. S. Kim NRC Project Manager, Millstone U. S. Nuclear Regulatory Commission, One White Flint North - Mail Stop 08 C2A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127
Serial No 12-485 MPS3 TSTF-510 - SG Program ATTACHMENT 1 Description and Assessment Dominion Nuclear Connecticut, Inc.
Millstone Power Station Unit 3
Serial No 12-485 MPS3 TSTF-510 - SG Program, Page 1 of 5 DESCRIPTION AND ASSESSMENT
1.0 DESCRIPTION
The proposed change revises Technical Specification (TS) 6.8.4.g, "Steam Generator (SG) Program" and 6.9.1.7, "Steam Generator Tube Inspection Report." The proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications. For consistency, additional administrative changes are being made to TS 3.4.5 "Steam Generator Tube Integrity."
The proposed license amendment is consistent with Technical Specification Task Force (TSTF) traveler, TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
2.0 ASSESSMENT
2.1 Applicability of Published Safety Evaluation Dominion Nuclear Connecticut, Inc. (DNC) has reviewed TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection,"
(ADAMS Accession No. MLI 10610350) and the model safety evaluation dated October 19, 2011 (ADAMS Accession No. ML112101513) as identified in the Federal Register Notice of Availability, dated October 27, 2011 (76 FR 66763).
As described in the subsequent paragraphs, DNC has concluded that the justifications presented in TSTF-510 and the model safety evaluation prepared by the Nuclear Regulatory Commission (NRC) staff are applicable to Millstone Power Station Unit 3 (MPS3) and justify this amendment for incorporation of the changes to the MPS3 TSs.
2.2 Optional Changes and Variations DNC is not proposing any technical variations or deviations from the TS changes described in TSTF-510, Revision 2, or the applicable parts of the NRC staff's model safety evaluation.
However, DNC is proposing the following administrative variations from the TS changes described in TSTF-510, Revision 2.
The MPS3 TS numbering system is different than the Improved Technical Specifications (ITS) on which TSTF-510 was based. Specifically, the "Steam Generator (SG) Program" in the MPS3 TS is numbered 6.8.4.g rather than 5.5.9, the "Steam Generator Tube Integrity" TS is numbered 3.4.5 rather than 3.4.17, and the "Steam Generator Tube Inspection Report" is numbered 6.9.1.7 rather than 5.6.7.
These differences are administrative and do not affect the applicability of TSTF-51 0 to the MPS3 TSs.
Serial No 12-485 MPS3 TSTF-510 - SG Program, Page 2 of 5 In addition, the proposed change corrects an administrative inconsistency in TSTF-510, Paragraph d.2 of the Steam Generator Tube Inspection Program.
In Section 2.0, "Proposed Change," TSTF-510 states that references to "tube repair criteria" in Paragraph d.2 is revised to "tube plugging [or repair] criteria." However, in the following sentence in Paragraph d.2, this change was inadvertently omitted, "If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated" (emphasis added).
The underlined phrase should state "tube plugging [or repair] criteria," consistent with the other changes made in TSTF-510.
DNC is changing the phrase to "tube plugging criteria." This change is administrative and should not result in this application being removed from the Consolidated Line Item Improvement Process.
This administrative error was identified in a February NRC-TSTF meeting and documented in a letter from the TSTF to the NRC dated March 28, 2012 (TSTF letter No.
12-09).
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Determination Dominion Nuclear Connecticut, Inc. (DNC) requests adoption of an approved change to the standard technical specifications (STS) into the plant specific technical specifications (TS) for Millstone Power Station Unit 3 (MPS3), to revise TS 6.8.4.g, "Steam Generator (SG) Program," TS 6.9.1.7, "Steam Generator Tube Inspection Report," and TS 3.4.5, "Steam Generator Tube Integrity," to address inspection periods and other administrative changes and clarifications.
As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that are analyzed as part of a plant's licensing basis.
The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident
Serial No 12-485 MPS3 TSTF-510 - SG Program, Page 3 of 5 analysis.
The proposed change will not cause the consequences of a SGTR to exceed those assumptions.
The proposed change to reporting requirements and clarifications of the existing requirements have no affect on the probability or consequences of SGTR.
Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the SG Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation.
In addition, the proposed change does not impact any other plant system or component.
Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory.
As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.
Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.
Serial No 12-485 MPS3 TSTF-510 - SG Program, Page 4 of 5 Based on the above, DNC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
3.2 Applicable Regulatory Requirements/Criteria During the initial plant licensing of MPS3, it was demonstrated that the design of the reactor coolant pressure boundary (RCPB) met the regulatory requirements in place at that time. This section evaluates the design bases of MPS3 as measured against the NRC General Design Criteria (GDC) for Nuclear Power Plants, Appendix A to 10 CFR 50, as amended through October 27, 1978.
The following information demonstrates compliance with GDC 14, 15, 30, 31, and 32 of 10 CFR 50, Appendix A. Specifically, Section 3.1.1 of the Updated Final Safety Analysis Report (UFSAR) discusses the design of the station relative to the design criteria published in 1978. The GDC state that the RCPB shall have "an extremely low probability of abnormal leakage... and gross rupture" (GDC 14/UFSAR 3.1.2.14), "shall be designed with sufficient margin" (GDC 15/UFSAR 3.1.2.15 and GDC 31/UFSAR 3.1.2.31), shall be of "the highest quality standards practical" (GDC 30/UFSAR 3.1.2.30), and shall be designed to permit "periodic inspection and testing..
. to assess..
. structural and leak tight integrity" (GDC 32/UFSAR 3.1.2.32). Structural integrity refers to maintaining adequate margins against burst, and collapse of the SG tubing. There are no changes to the SG design that impact these GDC.
The TS plugging limits ensure that tubes accepted for continued service will retain adequate structural and leakage integrity during normal operating, transient, and postulated accident conditions. The RCPB is designed, fabricated, and constructed so as to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime.
RCPB components have provisions for the inspection, testing, and surveillance of critical areas by appropriate means to assess the structural and leaktight integrity of the RCPB components during their service lifetime. Structural integrity refers to maintaining adequate margins against burst, and collapse of the SG tubing. Leakage integrity refers to limiting primary-to-secondary leakage during all plant conditions to within acceptable limits.
10 CFR 50, Appendix B, establishes quality assurance requirements for the design, construction, and operation of safety related components. The pertinent requirements of this appendix apply to all activities affecting the safety related functions of these components. These requirements are described in Criteria IX, XI, and XVI of Appendix B and include control of special processes, inspection, testing, and corrective action.
Under 10 CFR 50.65, the Maintenance Rule, licensees classify SGs as risk significant components because they are relied upon to remain functional during and after design basis events.
SGs are to be monitored under 10 CFR 50.65(a)(2) against industry established performance criteria. Meeting the performance criteria of NEI 97-06, Revision 3, provides reasonable assurance that the SG tubing remains capable of fulfilling its specific safety function of maintaining the RCPB.
Serial No 12-485 MPS3 TSTF-510 - SG Program, Page 5 of 5 4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
Serial No 12-485 MPS3 TSTF-510 - SG Program Marked-up Technical Specifications Changes Dominion Nuclear Connecticut, Inc.
Millstone Power Station Unit 3
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 Steam Generator (SG) tube integrity shall be maintained.
AND All SG tubes satisfying the tube ropaicriteria shall be plugged in accordance with the Steam Generator P-rogrm APPLICABILITY:
MOD,,
2, 3, and 4.
ACT--ION.": _-..
N OTE - -
C*I N entry is allowed for each S_ tube.
- a.
o*one or more SG tubes satisfying the tub criteria and not plugged in acco ance with the Steam Generator Program:
)
1.
tube integrity of the affected tube(s) is maintained until the next ng outage or SG tube inspection within 7 days, and
- 2.
Plug the Kected tube(s) in accordance with the Steam Generator Program prior to ente g HOT SHUTDOWN following the next refueling outage or SG tube inspe on.
- b.
With required ACTION an associated completion time of ACTION a. not met or SG tube integrity not maintai d:
- 1.
Be in HOT STANDBY wi in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- 2.
Be in COLD SHUTDOWN wit
- 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with the Steam enerator Program.
4.4.5.2 Verify that each inspected SG tube that satisfies the tube m criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.
49 MILLSTONE - UNIT 3 3/4 4-14 Amendment No. Q.M-
ADMINISTRATIVE CONTROLS
- g.
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 p*i*':
- 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 a SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
- b.
Provisions for 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 dov
&A-Saii anticipated transients included in the design specificatio a~ndd design boasis acci~dents. Thi
-efhf*- "
sa ety" cý
ý
ý u s-ýnins urst under normal steady state full power operation primary-to-secondary pressure differential and a
- safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or a combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- 2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG MILLSTONE - UNIT 3 6-17a Amendment No. 2.3&
Oz*;L.*
7, 2011 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
Leakage is not to exceed 500 gpd per SG.
- 3. The operational LEAKAGE performance criterion is specified in RCS LCO 3.4.6.2, "Operational LEAKAGE."
- c.
Provisions for SG tu rerair criteria: Tubes found by inservice in am flaws with a depth equal to or exceeding 40%
plugging o the nominal tube wall thickness shall be plugged.
The following alternate tu
-4ieper criteria shall be applied as an alternative to the 40% depth-based criteria:
- 1. For Refueling Outage 14 and the subsequent operating cycle, tubes with service-induced flaws located greater than 15.2 inches below the top of the tubesheet do not require plugging.
Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.2 inches below the top or of the tubesheet shall be plugged upon detection.
MILLSTONE - UNIT 3 6-17b Amendment No. 2 3 8, 24 5,-249,-26 Oetob. q, g@11 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 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 plugging tube, from the tube-to-tubesheet weld at the tube inlet to the tu besheet weld at the tube outlet, and that may satisfy the applicable tu t or criteria. For Refueling Outage 14 and the subsequent operating cycle, portions of the tube below 15.2 inches below the top of the tubesheet are excluded from this requirement.
The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d. 1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A~assessment shall be performed to dete type and location of flaws to which the tubes may be degradation 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 SQ installation
- 2. Inspect 100% o the tubes at equential periods of 120, 90, and, ther fter, 60 fective full wer months.
e first sequen al per*d shall considered o begin after tte first inservic in ection the SGs. In ddition, inspe 50% of the tu s by t e refueli g outage ne est the midpoi t of the period nd the Insert A remainin 50% by the efueling outa nearest the en of the period.
o SG shall erate for mor than 48 effecti e full power onths or tw refueling ou ges (whichever s less) without being inspected.
- 3. If crack indications are found in portions of the SG tube not excluded above, then the next inspection for SG for the degradntioetalvafete egaato"se the crack indication shall affected and potentially
.not exceed 24 effective full power months or one refueling outage (whichever=o4e). If definitive information such as from examinati df a pulled tube, diagnostic non-destructive testing, or ineering evaluation indicates that a crack-like ic n Is not results in more frequent inspections MILLSTONE - UNIT 3 6-17c Amendment No. 69, 4-86, 2+-2, 238-243, 24524, 25
týJLLmhm T, 26 t ADMINISTRATIVE CONTROLS 6.9.1.6.c The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.
6.9.1.6.d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with TS 6.8.4.g, Steam Generator (SG)
Program. The report shall include:
- a.
The scope of inspections performed on each SQ
- b.
A6,*-,r gradation mechanisms found,
- c.
Nondestructive examination techniques utilized for each degradation mechanism,
- d.
Location, orientation (if linear), and measured sizes (if available) of service induced indications,
.e.
Number of tubes plugged during the inspection outage for each ae4,,degradation mechanism,
- f. *Tcta :;.-*~ran-d pe~reentage 4wfbes,,- plugged to, date,
- g.
e results of condition monitoring, including the results of tube pulls and in-situ e is ng, During efueling Outage 14 and the subsequent operating cycle, the primary to seconda LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKA to an individual SQ the entire primary to secondary LEAKAGE should be c nservatively assumed to be from one SG) during the cycle preceding
- h.
the inspectio which is the subject of the report, MILLSTONE - UNIT 3 6-21 Amendment No. 24, 40, 5O, 69, -04, 4-7;2 2-2 4,29, 23 ~
24 ~9. 2;2 The number and percentage of tubes plugged to date and the effective Plu~qqin~q in e~ach stea~m qenerator..............
cs.i.~jb~i -,1, ~
11 ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (Continued)
- j.
During Refueling Outage 14 and the subsequent operating cycle, the calculated I
accident induced leakage rate from the portion of the tubes below 15.2 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and
- k.
During Refueling Outage 14 and the subsequent operating cycle, the results of Jmonitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report.
6.10 Deleted.
6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA As provided in paragraph 20.1601 (c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:
MILLSTONE - UNIT 3 6-21a Amendment No. 2-3-,
2 4 5, 2 4 9, 2M
Serial No.12-485 MPS3 TSTF-510 - SG Program Insert A for MPS3 TS 6.8.4.q - SG Program (600TT tubes)
- 2.
After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.
a)
After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; b)
During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c)
During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.
Serial No 12-485 MPS3 TSTF-510 - SG Program Marked-up Technical Specifications Bases Changes (For Information Only)
Dominion Nuclear Connecticut, Inc.
Millstone Power Station Unit 3
LBDCR No.
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY LCO The LCO requires that steam generator (SG) tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the
.iopocriteria be plugged in accordance with the Steam Generator Program.
plugging During a SG inspection, any inspected tube that sa m Generator "fro.
criteria is removed from service by plugging. If a tube was determine to s criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.
A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.8.4.g, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.
The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.
MILLSTONE - UNIT 3 B 3/4 4-3 Amendment No.
LBDCR No. "-
5-M, M,*ay 25, 2006 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY (Continued) a.__ and a.22 ACTION a. applies if it is discovered e or more SG tubes exa ined in an inservice inspection satisfy the tube riteria but were not plugged in a ordance with the Steam Generator Program as required by SR 4.4.5.2. An evaluation of tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based meeting the SG performance criteria described in the Steam Generator Program. The SG 6ocriteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, ACTION b.
applies.
A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.
If the evaluation determines that the affected tube(s) have tube integrity, Required ACTION a.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tube(s). However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next-inspection is supported by the operational assessment.
b.1 and b.2 If the ACTIONS and associated Completion Times of ACTION a. are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.
MILLSTONE - UNIT 3 B 3/4 4-3b Amendment No.
LBDCR No.
,,-M,,3-ý REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY (Continued)
SURVEILLANCE REQUIREMENTS TS 4.4.5.1 During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.
During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.
p The Steam Generator Program determines the scope of the inspecti n e metho ed to determine whether the tubes contain flaws satisfying the tube rpr criteria. Insp J n ope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a functio exis ing and potential degradation locations. The Steam Generator Program also specifies
- e. spect n methods to be used to find potential degradation. Inspection methods are a unc on of degradation morphology, non-destructive examination (NDE) technique apab' ities, and inspection locations.
The Steam Generator Program defines the Frequency of TS 4.4.5.. The F quency is det rmined by the operational assessment and other limits in the SG examj ation gu' elines (Referen e 6).
The Steam Generator Program uses information on existing gradatio s and growth rate; to determine an inspection Frequency that provides reasonab assuranc that the tubing wil meet the SG performance criteria at the next scheduled inspe ion. In ad tion, Specification 6 8.4.g contains prescriptive-requirements concerning inspec ion intervals o provide added assur nce that the SG performance criteria will be met betwe scheduledj *spections.
TS 4.4.5.2 During a SG inspection, any inspected tu that satisfies t Steam Generator Program erew criteria is removed from service by pl ging. The tube criteria delineated in Specification 6.8.4.g are intended to ensure that t es accepted for continued service satisfy the SG performance criteria with allowa for error in the flaw size measurement and for future flaw growth. In addition, the tube op@-ip criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational MILLSTONE - UNIT 3 B 3/4 4-3c Amendment No.
I
LBDCR No. 46-MP2 00 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY (Continued) plugging assessments to verify that the tubes remaining in service will continue t eet the SG performance criteria.
The Frequency of prior to entering MODE 4 following a SG i ection ensures that the Surveillance has been completed and all tubes meeting the Neeii criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.
BACKGROUND SG tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.1.1, "STARTUP and POWER OPERATION," LCO 3.4.1.2, "HOT STANDBY,"
LCO 3.4.1.3, "HOT SHUTDOWN," and LCO 3.4.1.4.1, "COLD SHUTDOWN - Loops Filled."
SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.
SG tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.
Specification 6.8.4.g., "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.8.4.g., tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 6.8.4.g. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.
The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Reference 1).
MILLSTONE - UNIT 3 B 3/4 4-3d Amendment No.
Serial No.12-485 MPS3 TSTF-510 - SG Program Insert B for MPS3 TS Bases 4.4.5.1 - SG Tube Integrity Surveillance Requirements If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 6.8.4.g until subsequent inspections support extending the inspection interval.