NL-06-089, Reply to Request for Additional Information Regarding Proposed License Amendment for Adoption of TSTF 449

From kanterella
Jump to navigation Jump to search

Reply to Request for Additional Information Regarding Proposed License Amendment for Adoption of TSTF 449
ML062560048
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 08/30/2006
From: Dacimo F
Entergy Nuclear Northeast
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-06-089, TAC MD2178, TAC MD2179
Download: ML062560048 (33)
v • d • e


Text

Entergy Nuclear Northeast Indian Point Energy Center wwREnterWgm 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel 914 734 6700 Fred Dacimo Site Vice President Administration August 30, 2006 Re: Indian Point Units 2 and 3 Dockets 50-247 and 50-286 NL-06-089 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington. DC 20555-0001

SUBJECT:

Reply to Request for Additional Information Regarding Proposed License Amendment for Adoption of TSTF 449 (TAC MD2178 1 2179)

REFERENCES:

1. NRC letter to Entergy dated July 12, 2006; "Request for Additional Information Regarding Amendment Application to Revise Technical Specifications".
2. Entergy letter NL-06-063 to NRC dated May 31, 2006; "License Amendment Request for Adoption of TSTF 449 Regarding Steam Generator Tube Integrity".

Dear Sir or Madam:

Entergy Nuclear Operations, Inc (Entergy) is providing, in Attachment One, a response to the request for additional information (Reference 1) regarding a proposed license amendment (Reference 2) for the adoption of TSTF 449, "Steam Generator Tube Integrity".

Responses to the request for additional information are provided in Attachment One and revised markups for the affected Technical Specification and Bases pages are provided in Attachment Two. The information provided in this response does not change the conclusions of the No Significant Hazards Evaluation previously provided by Entergy in Reference 2.

There are no new commitments identified in this submittal. If you have any questions or ADD (

NL-06-089 Dockets 50-247 and 50-286 Page 2 of 2 require additional information, please contact Mr. Patric W. Conroy, IPEC Licensing Manager at (914) 734-6668.

_yerely, Fred R. Dacimo Site Vice President Indian Point Energy Center cc:

Mr. John P. Boska, Senior Project Manager, NRC NRR DORL Mr. Samuel J. Collins, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 2 NRC Resident Inspector's Office, Indian Point 3 Mr. Peter R. Smith, NYSERDA Mr. Paul Eddy, NYS Department of Public Service

ATTACHMENT ONE TO NL-06-089 REPLY TO REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED LICENSE AMENDMENT FOR ADOPTION OF TSTF 449 (STEAM GENERATOR TUBE INTEGRITY)

ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNITS NO. 2 and 3 DOCKETS 50-247 and 50-286

NL-06-089 Attachment One Dockets 50-247 and 50-286 Page 1 of 4 Question 1:

In proposed TS 5.5.7.a for Unit 2 and TS 5.5.8.a for Unit 3, you indicated that condition monitoring will be performed when steam generator tubes "are inspected, plugged, to confirm" that the performance criteria are being met. Please discuss your plans to correct this apparent typographical error. The phrase should read: "are inspected or plugged."

Enterqy Reply:

Entergy has corrected the affected pages (IP2 Insert 5.5.7 and IP3 Insert 5.5.8) to read "are inspected or plugged" (IP2 Insert 5.5.7 and IP3 Insert 5.5.8) and replacement markups are provided in Attachment Two.

Question 2:

In proposed TS 3.4.17 for Unit 2 and 3, the second limiting condition for operation contains the phrase "[or repaired]" whereas all other references to tube repair have been crossed out. Since repair is not authorized for Unit 2 or 3, please discuss your plans for removing this phrase.

Enterqy Reply:

Entergy has corrected the affected pages (IP2 3.4.17-1 and IP3 3.4.17-1) by removing the phrase "[or repaired]" and replacement markups are provided in Attachment Two.

Question 3:

In proposed TS 3.4.13, the added text for Action B is not clear. Please confirm that the proposed addition is "OR Primary to secondary leakage not within limit."

Entergy Reply:

This is to confirm that the proposed addition is as stated in the Question. A clearer markup page (IP3 3.4.13-1) is provided in Attachment Two.

Question 4:

The proposed wording for the last sentence in TS 5.5.8.b.2 for Unit 3 is not clear. It currently reads: "Leakage is not to exceed 0.3 gpm [gallons per minute] per SG not to exceed 1 gpm total leakage." Please discuss your plans to clarify this sentence in both your proposed TS and in your Bases (page B 3.4.17-3). For example, "Leakage is not to exceed 0.3 gpm per SG and 1 gpm through all SGs."

NL-06-089 Attachment One Dockets 50-247 and 50-286 Page 2 of 4 Entergy Reply:

Entergy has revised the affected pages (IP3 Insert 5.5.8 and B 3.4.17-3) to read "Leakage is not to exceed 0.3 gpm per SG and 1 gpm through all SGs." and replacement markups are provided in Attachment Two.

Question 5:

In your proposed Bases, the reference numbers do not appear correct. Please discuss your plans to correct the reference numbers in the Bases. For example:

Insert B 3.4.13 B for Unit 2: Reference 4 in the first paragraph should be Reference 5.

Insert B 3.4.13 D (WOG) for Unit 2: Reference 5 in the first and third paragraph should be Reference 6.

Insert B 3.4.13 B for Unit 3: Reference 4 in the first paragraph should be Reference 3.

Insert B 3.4.13 D (WOG) for Unit 3: Reference 5 in the first and third paragraph should be Reference 4.

Entergv Reply:

Entergy has reviewed the reference numbering used in the original markup package and corrected pages are provided in Attachment Two (1P2 Bases 3.4.13 Insert Pages 1 and 2; IP3 Bases 3.4.13 Insert Pages 1 and 2).

Question 6:

In the second paragraph on proposed TS Bases page B 3.4.17-2 for Unit 2, there is reference to General Design Criterion (GDC) 19 and Title 10 of the Code of FederalRegulations (10 CFR)

Part 100. However, on page B 3.4.13-2, there is reference to a Westinghouse report and NRC Regulatory Guide 1.183. Please confirm that 10 CFR 50.67 is not applicable to your current Unit 2 licensing basis dose assessments.

Similarly, on proposed TS Bases page B 3.4.13-2 for Unit 3, the dose consequences were indicated to be within the limits of 10 CFR 50.67 and the NRC staff approved licensing bases whereas on page B 3.4.17-2, the dose consequences are indicated to be within the limits of GDC1 9 and 10 CFR 100, or the NRC-approved licensing basis. Please clarify this apparent discrepancy.

NL-06-089 Attachment One Dockets 50-247 and 50-286 Page 3 of 4 Entergy Reply:

10 CFR 50.67 and Regulatory Guide 1.183 are part of the licensing basis of both Indian Point Unit 2 (License Amendments 211 and 241) and Indian Point Unit 3 (License Amendment 224).

Entergy has corrected the affected pages (IP2 B 3.4.17-2 and -7; IP3 B 3.4.17-2 and -7) and replacement markups are provided in Attachment Two.

Question 7:

In your current Unit 3 safety analyses for reactor coolant system operational leakage (refer to page B 3.4.13-2), there is a statement that the 1 gallon per minute (gpm) primary to secondary leakage is relatively inconsequential. You have proposed to change the 1 gpm to 0.9 gpm.

Please discuss the basis for revising this limit. In addition, confirm that the 0.9 gpm is consistent with your current licensing basis for Unit 3.

Enter-gy Reply:

The proposed change from 1 gpm to 0.9 gpm was not intended to be a change in the existing 1 gpm limit. The safety analysis, for SGTR discussed on Indian Point Unit 3 Bases page B.3.4.13-2 assumes 0.9 gpm primary to secondary leakage in the intact steam generators. The 0.1 gpm leakage remaining in the 1.0 gpm limit is assumed for the faulted steam generator but excluded from the model as insignificant compared to the leakage from the ruptured tube as allowed by Regulatory Guide 1.183. Entergy has deleted the proposed change regarding 0.9 gpm and the affected replacement page (IP3 B 3.4.13-2) is provided in Attachment Two.

Question 8:

On page B 3.4.13-2 of the Unit 3 TS Bases, there is a statement that the safety analysis for the steam line break accident assumes 432 gallons per day primary to secondary leakage through the affected steam generator. This statement appears to be incomplete. Please confirm that your current licensing basis steam line break accident analysis not only assumes 432 gallons per day leakage but also that the total leakage from all steam generators is limited to 1 gpm. If so, discuss your plans to modify your TS Bases. If this is not the case, please confirm the adequacy of insert B 3.4.13A for Unit 3, which indicates that leakage is limited to 1 gpm from all steam generators.

Enter-qy Reply:

The current licensing basis steam line break accident analysis limit assumes 432 gallons per day leakage through the affected steam generator and that the total leakage from all steam generators is limited to 1 gpm. The 432 gallons per day leakage is equivalent to 0.3 gpm.

Entergy has revised the affected page (B.3.4.13-2) and the replacement markup is provided in Attachment Two.

NL-06-089 Attachment One Dockets 50-247 and 50-286 Page 4 of 4 Question 9:

In the Unit 3 Bases for Surveillance Requirement (SR) 3.4.13.1 (page B 3.4.13-5), you indicated that the surveillance is modified by two notes. The next statement then starts with "Therefore, this SR....". TSTF-449 indicated that this should have been modified to indicate that "Note 1 states...". Please discuss your plans to modify the Bases to be consistent with TSTF-449.

Entergy Reply:

Entergy has corrected the affected page (IP3 3.4.13-5) to be consistent with TSTF 449, and the replacement markup is provided in Attachment Two.

Question 10:

Several of the marked-up TS pages you submitted do not match the previously NRC issued pages. For Unit 2, pages ii, iv, and 1.1-3 were issued by the NRC as Amendment No. 238. For Unit 3, page ii was issued by the NRC as Amendment No. 205, and page iv as Amendment No.

207. Please revise the marked-up pages to show the changes from the NRC issued pages.

Entergy Reply:

Attachment Two includes new markup pages for the IP2 and IP3 Table of Contents. In addition to the changes required to reflect TSTF 449, these markup pages incorporate changes from other license amendments previously approved by NRC, and reflect minor reformatting to improve layout and readability.

ATTACHMENT TWO TO NL-06-089 REPLACEMENT MARKUP PAGES FOR PROPOSED LICENSE AMENDMENT FOR ADOPTION OF TSTF 449 (STEAM GENERATOR TUBE INTEGRITY)

NOTE: The following 25 pages, which address this NRC RAI, replace corresponding pages previously provided in Entergy letter NL-06-063, dated May 31, 2006.

The remaining markup pages from NL-06-063 are not affected by this RAI reply.

IP2 IP3 Tech Spec Bases Tech Spec Bases RAI Tof C i- iv Tof C i- v 10 3.4.13-1 3 3.4.17-1 3.4.17-1 2 Insert 5.5.7 Insert 5.5.8 1 and 4 B 3.4.13-2 7 and 8 B 3.4.13-5 9 Insert B 3.4.13 Insert B 3.4.13 Page 1 of 2 Page 1 of 2 5 Insert B 3.4.13 Insert B 3.4.13 Page 2 of 2 Page 2 of 2 5 B 3.4.17-2 B 3.4.17-2 6 B 3.4.17-3 4 B 3.4.17-7 B 3.4.17-7 6 ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNITS NO. 2 and 3 DOCKETS 50-247 and 50-286

Facility Operating License No. DPR-26 Appendix A - Technical Specifications TABLE OF CONTENTS 1.0. USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.2 Reactor Coolant System Pressure SL 2.2. Safety Limit Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.1 L 3.1.1 Si ýTDQWN MARG IN (SD 3.1.2 Core Reactivity 3.1.3 Moderator Temperature Coefficient (MTC) 3.1.4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.7 Rod Position Indication

>3.1.8 H CS TESTS Exceptions - MODE 2 3.2 S1O PO E DIS TRIB T IO LIMIt 3.2.1 Flx Cannl Fc or F(Z))

Hea Ht 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNH) 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Constant Axial Offset Control (CAOC) Methodology) 1* .3.2.4 G QUADRANT POWERTILT RATIO (QPTR)

  • -3.3 INSTRUMENTATION 10L 3.3.1 Reactor Pr ection System (RPS) Instrumentation 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation 3.3.3 Post Accident Monitoring (PAM) Instrumentation 3.3.4 3.3.5 Remote Shutdown 3.3.6 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation Pý >3.3.7 Containment Purge System and Pressure Relief Line Isolation Instrumentation br~p- "*3.4 Control Room Ventilation System CRVS)Actuation Instrumentation RE CTOR COO- T SYSTE*M RC-  !'.D 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.4.2 RCS Minimum Temperature for Criticality INDIAN POINT 2 i Amendment No. 238

Facility Operating License No. DPR-26 Appendix A - Technical Specifications TABLE OF CONTENTS 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.4 RCS Loops - MODES 1 and 2 3.4.5 RCS Loops - MODE 3 3.4.6 RCS Loops - MODE 4 3.4.7 RCS Loops - MODE 5, Loops Filled 1'1JlSt,-- q-x*-r 3.4.8 RCS Loops - MODE 5, Loops Not Filled t" ) "

3.4.9 Pressurizer 3.4.10 Pressurizer Safety Valves 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) 3.4.12 Low Temperature Overpressure Protection (LTOP) 3.4.13 RCS Operational LEAKAGE 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage 3.4.15 RCS Leakage Detection Instrumentation I ' .3.4.16 RCS Specific Activity spc. 3.5 E NCY CORE COOLING SYSTEM (ECCS) DLD 3.5.1 Accumulators 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown

\ >.3.5.4 Refuelin Water Stora.e Tank (RWST) 3.6 CONTAIMENT SYSTEMS O 3.6.1 on ainment 3.6.2 Containment Air Locks 3.6.3 Containment Isolation Valves 3.6.4 Containment Pressure 3.6.5 Containment Air Temperature 3.6.6 Containment Spray System and Containment Fan Cooler Unit (FCU)

_System 3.6.7 Recirculation pH Control System 1o+/- _

3.6.8 H drogen Recombiners -

3.6.9 Isolation Valve Seal Water (IVSW) System

>3.6.10 Weld Channel and Penetration Pressurization System (WC&PPS)

.br51< 3.7 3.7.1 Main Steam Safety Valves (MSSVs) 3.7.2 Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs) 3.7.3 Main Feedwater Isolation 3.7.4 Atmospheric Dump Valves (ADVs) 3.7.5 Auxiliary Feedwater (AFW) System 3.7.6 Condensate Storage Tank (CST) 3.7.7 Component Cooling Water (CCW) System 3.7.8 Service Water System (SWS) 3.7.9 Ultimate Heat Sink (UHS) 3.7.10 Control Room Ventilation System (CRVS)

INDIAN POINT 2 ii Amendment No. 238

Facility Operating License No. DPR-26 Appendix A - Technical Specifications TABLE OF CONTENTS 3.7.11 Spent Fuel Pit Water Level 3.7.12 Spent Fuel Pit Boron Concentration 3.7.13 Spent Fuel Pit Storage

>.3.7.14 Seconda Specific Activity 3.8 ELECTRIC PqOWER SYSTEMSý &LD 3.8.1 AC Sources - Opera ing 3.8.2 AC Sources - Shutdown 3.8.3 Diesel Fuel Oil and Starting Air 3.8.4 DC Sources - Operating 3.8.5 DC Sources - Shutdown 3.8.6 Battery Parameters 3.8.7 Inverters - Operating

  • 3.8.8 Inverters - Shutdown 3.8.9 Distribution Systems - Operating

>istribution S stems - Shutdown 3.9 REFUELING OPERATI 130 LO 3.9.1 Boron Concentration 3.9.2 Nuclear Instrumentation 3.9.3 Containment Penetrations 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level 3.9.6 Refueling Cavity Water Level 4.0 DESIGN FEATURES 4.1 Site Location 4.2 Reactor Core 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.2.1 Onsite and Offsite Organizations 5.2.2 Unit Staff 5.3 Unit Staff Qualifications 5.4 Procedures 5.5 Programs And Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) 5.5.2 Primary Coolant Sources Outside Containment 5.5.3 Radioactive Effluent Controls Program 5.5.4 Component Cyclic or Transient Limit 5.5.5 Reactor Coolant Pump Flywheel Inspection Program INDIAN POINT 2 iii Amendment No. 238

Facility Operating License No. DPR-26 Appendix A - Technical Specifications TABLE OF CONTENTS 5.5.6 Inservice Testing Program 5.5.7 Steam Generator (SG) Tube Surveillance Program 5.5.8 Secondary Water Chemistry Program 5.5.9 Ventilation Filter Testing Program (VFTP) 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.11 Diesel Fuel Oil Testing Program 5.5.12 Technical Specification (TS) Bases Control Program 5.5.13 Safety Function Determination Program (SFDP) 5.5.14 Containment Leakage Rate Testing Program 5.5.15 Battery Monitoring and Maintenance Program 5.6 Reporting Requirements 5.6.1 Occupational Radiation Exposure Report 5.6.2 Annual Raýioloagical Environmental Operating Report 5.6.3 -RdocieEffluent Release Report Nr.

5.6.4 M 5.6.5 CORE OPERATING LIMITS REPORT (COLR) 5.6.6 Post Accident Monitoring Report 5.6.7 Steam Generator Tube Inspection Report INDIAN POINT 2 iv Amendment No. 238

SG Tube Integrity

'1 3.4 REAC;TC )R COOLANT SYSTEM (RCS)

Steam Generator (SG) Tube Integrity LCO 9" SG tube integrity shall be maintained.

3'I' AND

[ -rIAIejin satisfying the All SG tubesaccordance with the repair tube Steamcriteria Generator be plugged*J shall Program.

fJ APPLICABILITY: MODES 1,2, 3, and 4.

ACTIONS K il-r!

I-

--- E- - - - - --- - - - - - -- - - - - - - - - -

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 JG.. a. in refueling outage or SG accordance with the tube inspection.

Steam Generator Program. AND A.2 Plug rprjthe affected Prior to entering tube(s in accordance with MODE 4 following the the Steam Generator next refueling outage Program. or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

vInd icsiPD41 V 3.Lj.l~-f I r L~~i AnLKd~~{ N0.

TSTF-449, Rev. 4 INSERT 5.5* < tz 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 foundc 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 S jIof tubes. Condition monitoring assessments shall be conducted during each

' age ring which the SG tubes are inspectecdoplugged,4,,r MID& !1o confirm that the performance criteria are being met. - 5 k '

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 a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakaW- rate for all SGs and.leakage rate for an individual SG. Leakage is not to

-exceed*[.1l.gprh]jjper SG [,fixcep t.fr spe*f types 'of deje~datioý,at spedfi 1 icatlods asri ec *pmgraphp,,,d De = the Ste,afmGeKerator, ProgrqE4 ----

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE.
c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal t9 or exceeding T400/c of the nominal tube wall thickness shall be plugged r r.

1

,a *r ,-. TSTF-449, Rev. 4 INSERT B 3.4.13 A 16oDo go Ib - Z1' 'J that primary to secondary _EAKAGE from all steam generators (SGs) isi(o0'e ga(or'/eY miKWu6 or increases tolfore galu* pe,'mYu]as a result of accident induced conditions. The LCO requirement to limit primary to seconda LEAKAGE through any one SG to less than or equal to 150 gallons per day is ic ntlyess t the conditions assumed in the safety analysis.

It INSERT B 3.4.13 B (IqL]I.i

d. Primary to Secondary LEAKAGE Through Any One SG m The limit of 150 gallons per day per SG is based on the operational LEAKAGE U performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref.I*. The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

INSERT B 3.4.13 C Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

INSERT B 3.4.13 D BWO This SR verifies that pri ary to secondary LEA GE is less tha or equal to 150 gallons per day through any one S . Satisfying the prim to secondary AKAGE limit ensures that e operational LEAKAG performance criterion n the Steam Ge erator Program is met. If t SR is not met, complian'e with LCO 3.4.17, "S am Generator T be Integrity," should be ev uated.

The 150 gallons pe day limit is measured t room temper re as described in Refere ce 5.

The operational L KAGE rate limit app

  • s to LEAKAGE rough any one SG. If it i not practical to assi the LEAKAGE to an
  • dividual SG, all e primary to secondary KAGE should be cons rvatively assumed to from one SGr The Surveill ce is modified by a N e which states at the Surveillance is no equired to be performed til 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after esta ishment of ste y state operation. For S primary to secondary EAKAGE determinati n, steady state' defined as stable RCS ressure, tempera re, power level, press izer and make tank levels, makeup a letdown, and RCP seal inj tion and return flows.

The urveillance Frequency f 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a easonable interval to tr d primary to secondary L GE and recognizes e importance early leakage detectio in the prevention of ac The primary to condary LEA GE is determined usi continuous process mdents.

r diation monitors or radi hemnical grab ampling in accordanc ith the EPRI guidelines (Ref.

N 1

INSERT B 3.4.13 Dý iŽ1~T~3A3D [~74]TSTF-449,Rev. 4 This SR verifies that primary to sec ndary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying th primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performaT.r.t .rrion in the Steam Generator Program is met. Ifthis SR is not met, compliance with LCO -'-Steam Generator Tube Integrity," should be evaluated.

The 150 gallons per day limit is measured at room temperature as described in Reference%7* r;G The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not "---

practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, 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.

The Surveillance Frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of p accidents. The primary to secondary LEAKAGE is determined using continuous process H 1.J radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. (d\

This SR verifies that pri ry to secondary LEA GE is less or equal to 150 gallons per y through any one SG. atisfying the primary t secondary LEAKAGE limit ensures tha e operational LEAKA performance criterio in the Steam Generator Program is m . If this SR is not met, compl* nce with LCO 3.4.18," team Generator Tube Integrity," shoul e evaluated.

The 150 gallon per day limit is meas d at room temperature as described i eference 5.

The operatio I LEAKAGE rate limi pplies to LEAKAGE through any one . If it is not practical t ssign the LEAKAGE an individual SG, all the primary to s ondary LEAKAGE should b conservatively asu** ed to be from one SG.

The urveillance is modif by a Note which states that the Surv i ance is not required t e orme until 12 hou after establishment of steady state op ation. For RCS prima to econdary LEAKAG etermination, steady state is defined stable RCS pressure temperature, pow level, pressurizer and makeup tank le Is, makeup and letdo and RCP seal injection a return flows.

The Surve nce Frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reas able interval to trend p( ary to secondary in t prevention of LEAK and recognizes the importance of e y leakage detection accid'leA ts. The primary to secondary LEAKA is determined using c tinuous process ra tion monitors or radiochemical grab s pling in accordance wi the EPRI guidelines (Ref.

INSERT B 3.4.13 E NEI 97-06, "Steam Generator Program Guidelines."

EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

2

I TSTF 449., Rew.-4 1 SG Tube Integritv.

I BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3A.13, "RCS Operational LEAKAGE,' plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid i is Majrk,fiyp released to isr discMMMarg the atmosphere via,.safety valves .Tm o7MemAiryco~aensns -*;t j/'* :- '.; *¢ l The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) Inthese analyses, the steam discharge to the atmosphere Is based on the total primary to secondary LEAKAGE from all SGs oiafp, ml .or is assumed to increase tol Lgop§E .660 (00 aIlO.'s O as a result of accident induced conditions. For accidents that doi*tlo.,,

'not involve fuel damage, the primary coolant activity level of DOSE per EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, URCS Is,,I Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released Lu o~fpI~LC4_*~l )ILfrom thedamaged fuel. The dose consequences of these events are o CPa S,0. w3 thin o&oMC 1(Ref.. 10CFR 100Ref. 3) rthe-QR, (R, A R roy atlic nsin4 basis (e.g., alsmall f/action these mits)./"

  • A,oII-i B ; (.j ** Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).6-LCO The LCO requires that SG tube integrity be maintained. The LCO also reuires that all SG tubes that satisfy the repair criteria be pluggedwi jr,6&9dyj In accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is V4AJre~or removed from service by plugging. If a tube was determined to sateis repair criteria but was not plugged_ repa,&d] the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as th r length of the tube, including the tube wall r 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 S erformance criteria.

The SG performance criteria are defined in Specificatiorf "Steam Generator Program, and describe acceptable SG tube pfe ormance.

The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

I I -- HIa All'.

^T&

WW%.RAJ Sr -1 I44.4.,W-2 I,(--' I --

I Itvv.

.1-lr%.^;:Il <--I

SG Tube Integrity REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10GF, 50 AC.o ndixo.

..... 0 G7.1

4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.
5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.
6. EPRI, "Pressurized Water Reactor Steam Generator Examination flc Ae *v-t in* *f ,.,{ , ,,Jty' cý72c ')

e3.ti1-I I.

FACILITY OPERATING LICENSE No DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS 1.6 USE ANDAPP)ICATION 1.1 De i ions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency LIMITS SLs) ( ,SAFETY 2.0 2.1 Safety Limits 2.2 Safety Limit Violations 3.0 LIIIGCNITION FOR OPERATIONS (LCO) APPLICABILITYt L 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.1 -

3.1.1 SHUTDOWN MARGIN 3.1.2 Core Reactivity 3.1.3 Moderator Temperature Coefficient (MTC) 3.1.4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.7 Rod Position Indication 3.1.8 PHYSICS TESTS Exceptions-MODE 2 P DISTRIBUTION L GoL-D 3.2 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FN'H) 3.2.3 AXIAL FLUX DIFFERENCE (AFD) 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

UNMENTATION q L 3.3 3.3.1 Reactor Protection System (RPS) Instrumentation 3.3.2 Engineered Safety Feature Actuation System (ESFAS)

Instrumentation 3.3.3 Post Accident Monitoring (PAM) Instrumentation 3.3.4 Remote Shutdown 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation 3.3.6 Containment Purge System and Pressure Relief Line Isolation Instrumentation (continued)

INDIAN POINT 3 i AMENDMENT 205

FACILITY OPERATING LICENSE No DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS ConTRUENAIol R tilaoný

ýoom 3.3.7 Control Room Ventilation (CRVS) Actuation Instrumentation 3.3.8 Fuel Storage Building Emergency Ventilation System (FSBEVS) 4ctuation Instrumentation 3.4 R SYSTEM (RCS) 13OL-D 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.4.2 RCS Minimum Temperature for Criticality 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.4 RCS Loops-MODES 1 and 2 3.4.5 RCS Loops-MODE 3 3.4.6 RCS Loops-MODE 4 3.4.7 RCS Loops-MODE 5, Loops Filled 3.4.8 RCS Loops-MODE 5, Loops Not Filled 3.4.9 Pressurizer 3.4.10 Pressurizer Safety Valves 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) 3.4.12 Low Temperature Overpressure Protection (LTOP) 3.4.13 RCS Operational LEAKAGE 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage 3.4.15 RCS Leakage Detection Instrumentation L4. II 3.4.16 RCS Specific Activity (SQCO~

3.5 EERGENCY CORE COOLNG!SYSTEMS(ECCS)

-Accumulators **

3.5.1 3.5.2 ECCS-Operating 3.5.3 ECCS -Shutdown 3.5.4 Refueling Water Storage Tank (RWST) 3.6 CONTAINMENT SYSTEMS 13OL-O 3.6.1 Cotanment 3.6.2 Containment Air Locks 3.6.3 Containment Isolation Valves 3.6.4 Containment Pressure 3.6.5 Containment Air Temperature (continued)

INDIAN POINT 3 ii AMENDMENT 205

FACILITY OPERATING LICENSE No DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS LL 3.6 CONTAINMENT SYSTEMS (continued) 3.6.6 Containment Spray System and Containment Fan Cooler System 3.6.7 Spray Additive System _

3.6.8 *Hydroen Recombiners-itje 3.6.9 Isolation Valve Seal Water (IVSW) System 3.6.10 Weld Channel and Penetration Pressurization System (WC & PPS) 3.7 PLANT YSTEMS l SotL.

3.7.1 Main Steam Safety Valves (MSSVs) 3.7.2 Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs) 3.7.3 Main Boiler Feedpump Discharge Valves (MBFPDVs), Main Feedwater Regulation Valves (MFRVs), Main Feedwater Inlet Isolation Valves (MFIIVs) and Main Feedwater Low Flow Bypass Valves 3.7.4 Atmospheric Dump Valves (ADVs) 3.7.5 Auxiliary Feedwater (AFW) System 3.7.6 Condensate Storage Tank (CST) 3.7.7 City Water (CW) 3.7.8 Component Cooling Water (CCW) System 3.7.9 Service Water (SW) System 3.7.10 Ultimate Heat Sink (UHS) 3.7.11 Control Room Ventilation System (CRVS) 3.7.12 Control Room Air Conditioning System (CRACS) 3.7.13 Fuel Storage Building Emergency Ventilation System (FSBEVS) 3.7.14 Spent Fuel Pit Water Level 3.7.15 Spent Fuel Pit Boron Concentration 3.7.16 Spent Fuel Assembly Storage 3.7.17 Secondary Specific Activity 3.8 LE I YSTM LZO 3.8.1 AC Sources-Operating 3.8.2 AC Sources-Shutdown 3.8.3 Diesel Fuel Oil and Starting Air 3.8.4 DC Sources-Operating 3.8.5 DC Sources-Shutdown (continued)

INDIAN POINT 3 iii AMENDMENT 207

FACILITY OPERATING LICENSE No DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS 3._*__.8 ELECTRICAL POWER SYSTEMS (continued)'

3.. Battery Cell Parameters-'*J "

3.8 . Inverters -Operating S 3.8.8 Inverters- Shutdown

"* 3.8.9 Distribution Systems-Operating 3.8.10 Distribution Systems-Shutdown 3.9 REFUELING OPERATIONS I.PL~.D 1

3 *.9.* -- Boron -oncentration 3.9.2 Nuclear Instrumentation 3.9.3 Containment Penetrations 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation-High Water Level

\3.9.5- Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level 396Refueling Cavity Water Level _ f-4.0 DEIG- EAUE S *L.

S 4.1 Site Location 4.2 Reactor Core 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 'DLD 5.1 Responsibility 5.2 Organization 5.3 Unit Staff Qualifications 5.4 Procedures 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) 5.5.2 Primary Coolant Sources Outside Containment 5.5.3 Post Accident Sampling 5.5.4 Radioactive Effluent Controls Program 5.5.5 Component Cyclic or Transient Limit 5.5.6 Reactor Coolant Pump Flywheel Inspection Program 5.5.7 Inservice Testing Program 5.5.8 Steam Generator (SG) T beSurv*eillance Program 5.5.9 Secondary Water Chemistry Program 5.5.10 Ventilation Filter Testing Program (VFTP)

(continued)

INDIAN POINT 3 iV AMENDMENT 207

FACILITY OPERATING LICENSE No DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS 5.0 ADMINISTRATIVE CONTROLS (continued) .-

5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program

?C, 5.5.12 Diesel Fuel Oil Testing Program LV 5.5.13 Technical Specification (TS) Bases Control Program 5.5.14 Safety Function Determination Program (SFDP) 5.5.15 Containment Leakage Rate Testing Program

'-7 - ZP.-

5.6 Reporting Requirments 5.6.1 ccuational Radiation Ex osure Report 3

5.6.2 Annual Ryadio1-gical Environmenta perating Report 5.6.3 Radioactive Effluent Release Report LOS _*

5.6.4 ti Repor -s 5.6.5 CORE OPERATING LIMITS REPORT (COLR) 5.6.6 NOT USED 5.6.7 Post Accident Monitoring Instrumentation (PAM) Report 5.6.8 Steam Generator Tube Inspection Report

.7 High Radiation Area-PcýL--QA - i INDIAN POINT 3 V AMENDMENT 207

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLNT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE:
b. 1 gpm unidentified LEAKAGE:
c. 10 gpm identified LEAKAGE:J Id./ 1antotalrima1ry/tsnda rhroh 11 ý 9dw*ef=at ( _.a n*rd /

[11---

gd [9'gallons

_j one - S4a per day primary to secondary LA 6 t

?( ) LEAKAGE through any APPLICABILITY: MODES 1. 2 . 3. and 4.

ACTIONS L A.

CONDITION R' LEAKAGE not within limits for reasons other A.1 REQUIRED ACTION Reduce LEAKAGE to within limits.

COMPLETION TIME 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> than pressure boundary toLEAKNIYLEK9 B. Required Action and B.1 Be in MOODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not ANM met.

8.2 Be in MOOE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Pressure boundary LEAKAGE exists.

.1roi- WIK- 1 INDIAN POINT 3 3.4 INDIAN POINT 3

.3.4Aedmn10 Amendment 205

SG Tube Integrty 3.4 REACTO R COOLANT SYSTEM (RCS)

StE.am Generator (SG) Tube Integrity LCO SG tube integrity shall be maintained.

AND

,ASG tubes satisfying the tube repair criteria shall be plugged*-

_. r]n accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2,3, and 4.

ACTIONS k

..... . ...-.-----... ------- ..----- f--IL I -------------------------------------------------------

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 f [, e [iein refueling outage or SG accordance with the tube inspection.

Steam Generator Program. AND A.2 Plug([tr the affected Prior to entering tube

  • accordance with MODE 4 following the the Steam Generator next refueling outage Program. or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

wee e~3T, F4~r- r >-

Afine-YAm4i No.

plw )

TSTF-449, Rev. 4 fCor TP 3 INSERT 5.59 1___

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 o tubes. Condition monitoring assessments shall be conducted during each outage'a-Tring which the SG tubes are inspectedjPluggedt4 .'paiFr]qJto confirm that the performance criteria are being met. ..9- --
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 a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment 0.3 3 M of tube integrity, those loads that do significantly affect burst or collapse shall be ppLr :ýC- cwfl 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.

I PrvM tihrotk oXil SG)s 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 SG. and.leakage rate for an individual SG. Leakage is not to co ret'l excee [1 ]pe-er $6 [,,gdepbgfor speifc types ef de~edatiopat spgeific t.\~/~ epafgraph.,-f e/.-Or e*9 the Sta Gerferator.,Prog-ml

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal tq or exceeding VI0°/4of the nominal tube wall thickness shall be plugged? re .

1

RCS Operational LEAKAGE B 3.4.13 BASES APPLICABLE SAFETY ANALYSES Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for events resulting in steam discharge to the atmos here assumes a ra e ffprmary o seco/dary YKAGE ?om 0.1/jpm to gpm as h inial 1133. :1-?. A /cond~tion./ I - - _ . V-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 FSAR (Ref. 2) analysis for SGTR assumes the contaminated secondary S ..'IS fluid is released via safety valves and atmospheric dump valves. The lo.s~uJr-lltog'*~ I1 gpm primary to secondary LEAKAGE is relatively inconsequential.

The SLB is more limiting for site radiatio releases. he safety analysis for the SLB accident assumesla range of7primary to secondary 4 is -t vo~k 4_* LEAKAGEas an initial condition. The dose consequences resulting from c the SLB-accident are well within the limits defined in 10 CFR 50.67 (S -S and the staff approved licensing basis (i.e., a small fraction of 4J aa*k fP these limits). *-

'V!3YP1Ak o c1 . The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36.

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting (continued)

INDIAN POINT 3 B.3.4.13 - 2 Revision 3

RCS Operational LEAKAGE B 3.4.13 BASES ACTIONS B1nd.2 (continued) reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 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 and gaskets is not pressure boundary LEAKAGE. 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 />. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full 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 further deterioration is much less likely.

SURVEILLANCE REQUIREMENTS SR 3.4.13.1 Verifying RCS LEAKAGE to be within the LCO 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. Pri ary tg secon y GE is so mea re pormace of CS ater ynventor balan in c9juncti with pflerf moni ~ring ithin ~he sec 'dary s earn blow sys The RCS water inventory balance must be met with the reacr at teady state operating conditions and near operating pressure.

this SR is not required to be performed in MODES 3 and 4 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation near operating pressure have been established.

The * ("rl cntiud (continued)

INDIAN POINT 3 B.3.4.13 - 5 Revision 3

TSTF-449, Rev. 4 INSERT B 3.4.13 A that primary to secondary LEAKAGE from all steam generators (SGs) is ne gallon per minut*

or increases to lne gallon per minute~as a result of accident induced coWhditions. The LCO U 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.

INSERT B3.4.13 B

d. Primary to Secondary LEAKAGE Throuch Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref.g. The VJ Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day.' The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

INSERT B 3.4.13 C Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

I

TSTF-449, Rev. 4 cfe2V-INSERT B 3.411 D This SR verifies that primary to s dary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying th4 primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performa rirl don in the Steam Generator Program is met. Ifthis SR is not met, compliance with LCO 4 Steam Generator Tube Integrity,* should be evaluated. X)z The 150 gallons per day limit Is esured at room temperature as described In Referencer,*-.rl The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. Ifit is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, 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.

The Surveillance Frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the Importance of early leakage detection in the prevention of :1, accidents. The primary to secondary LEAKAGE Is determined using continuous process radiation monitors or radiochemical grab sampling Inaccordance with the EPRI guidelines (Ref. Li.'

This SR verifies that p-- ry to secondary LEA GE is less or equal to 150 gallons per y through any one SG. tisfying the primary t econdary LEAKAGE limit ensures tha e operational LEAKA performance crite n the Steam Generator Program is m Ifthis SR is not met, compl ce with LCO 3.4.18, ea Generator Tube lntegrty, shoul evaluated.

The 150 gallon per day limit Is meas at room temperature as described I eference 5.

The operatio LEAKAGE rate 1im, pplies to LEAKAGE through any one . If it is not practical t ssign the LEAKAG an Individual SG, all the primary to s ndary LEAKAGE should conservatively assu to be from one SG.

Th urveillance is modif by a Note which states that the Surv ance is not required pformed until 12 hou after establishment of steady state op tion. For RCS prima to econdary LEAKAG determination, steady state is defined stable RCS pressure temperature, po level, pressurizer and makeup tank le Is,makeup and letdo, and RCP seal injection a retum flows.

The Surve ce Frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a rea able interval to trend p ary to secondary LEAKA and recognizes the importance of e leakage detectio in prevention of accid ts. The primary to secondary LEAKA is determined using c ; tinuous process ra tion monitors or radiochemical grab s piing in accordancew? the EPRI guidelines (Ref.

J 77 INSERT B 3.4.13 E NEI 97-06, "Steam Generator Program Guidelines.*

EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

2

"-F- 44-. Rev. 4 SGTube Igrty BASES 1831,14 APPUCABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE,' plus the leakage rate associated with a double-ended rupture of a single tube. The accidentanis for a SGTR assumes the contaminated secondary fluid is 19rI*dft released to the atmosphere via safety valvestatq2th (- .

Oýe Or 4MO5-bhx'oýh4t r ý.tA v1's The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (Le., they are assumed not to rupture.) In these analyses, the steam discharge to the

\

atmosphere is based on the total primary to secondary LEAKAGE from all SGs 011 gallon per minut.for is assumed to increase tog gallon per minutelas a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE OfI iLoe

_i1AS EQUIVALENT 1-131 Is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity, limits. For accidents that assume fuel damage, the primary coolant activity Is a function of the amount of activity released from the damaged fuel. The dose consequences of these evgnts are within"m of GI 19 (qef. 2), r CFR 100 (Ref. A)or the/NRC Iapr lice sing sis (e.,a Ifract of thea*limits)

I I Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires Ahat all SG tubes that satisfy the repair criteria be pluggedIR in accordance with the Steam Generator Program.

During an SG Inspection, any Inspected tube that satisfies the Steam Generator Program repair criteria isFIe-irIrlr removed from service by plugging. If a tb ibe was determined to satisfy the repair criteria but was not plugged rypE)rgdj 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[Vn rar# rpofs pfadd tolit 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 erformance criteria.

The SG performance criteria are defined in Specificatiorf "Steam Generator Program,* and describe acceptable SG tube pe oianca.

The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

'~

4

'LA ?oajf LB 3-1 98.4 i- J

I 4 ISTe-4*. -1 SG Tube Integrity BAsEs 3.

LCO (continued) 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.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other o3 ~ than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not excejedi[1 pm/

SG On S,, eept or specilficppes 9fdegiadati~onat/speofic Io,ati ~s//

Ip*'

V*he r t~h2NRa has*.pproed gr*tater,&ccidq~nt ir~duce..d leal6e, ITh~e L, accident induced leakage rate includes any primary to secondary VvoA~G.~ LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

'Lt.I- kýM6114Ae, l

IK110 JNGSag 1 13 0.4.20 a- I<J 11 Jý

  • r_ Irv A"1.

A ;l n -. 1 SG Tubengri REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. CFR 50 A .pnn.i,A, .,,.1..'.- lo CA* .

4.

5.

ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam I

Generator Tubes," August 1976.

6.

C EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

4 b-+

  • jio-L,- ýT %N, VC Iyy).

,, s Tu' ý -

jAl'A y

Rvct----PDvQ" ") :rL-d b~ut I 63A.R:4 I.

M:a.:4 pfl:F I

Retrieved from "https://kanterella.com/w/index.php?title=NL-06-089,_Reply_to_Request_for_Additional_Information_Regarding_Proposed_License_Amendment_for_Adoption_of_TSTF_449&oldid=2186487"