Text

License Amendment Request to Revise Technical Specifications to Adopt TSTF-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection Using the Consolidated Line Item Improvement Process
	ML14210A484
Person / Time
Site:	Salem
Issue date:	07/28/2014
From:	Jamila Perry; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LAR S14-02, LR- N14-0146
	Download: ML14210A484 (36)
	v • d • e

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 JUt 28'2014 LR-N14-0146 LAR S14-02 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 PSEG Nudea-r UJC 10 C FR 50.90

Subject:

License Amendment Request to Revise Technical Specifications to Adopt TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" Using the Consolidated Line Item Improvement Process Pursuant to 10 C FR 50.90, PSEG Nuclear LLC ( PSEG) hereby requests an amendment of the Technical Specifications (TS) for Salem Generating Station, Units 1 and 2.

The proposed amendment would modify TS requirements regarding steam generator tube inspections and reporting as described in TSTF-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection." provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides existing TS Bases

pages-marked-up-to-show-the-proposed-changes:-'fhe-Bas-es-m-arkop-p-ages-are-b-ein-g-p-roviaea'----

for information only.

In addition, PSEG requests to revise the Salem Generating Station Unit 2 Technical Specifications (TS) 6.8.4.i- "Steam Generator (SG) Program," 6.9.1.1 0- "Steam Generator Tube I nspection Report," and the Bases Section of 3/4.4.6 "Steam Generator (SG) Tube Integrity", to remove unnecessary information related to the original Salem Unit 2 Westinghouse steam generators.

The changes in this License Amendment Request (LAR) are not required to address an immediate safety concern; PSEG requests approval of this LAR in accordance with standard

LR-N14-0146 Page 2 10 CFR 50.90 NRC approval process and schedule. Once approved, the amendment will be implemented within 60 days from the date of issuance.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

If you have any questions or require additional information, please contact Ms. Tanya Timberman at 856-339-1426.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on JUl 28.2014 (Date)

Respectfully, cJd-r

Joh n F. Perry U

Site Vice President-Salem Generating Station Attachments:

1. Description and Assessment

2. Mark-up of Proposed Technical Specification Pages

3. Mark-up of Proposed Technical Specification Bases Pages cc:

Mr. W. Dean, Administrator, Region I, NRC Mr. J. Lamb, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Manager IV, NJBNE PSEG Corporate Commitment Tracking Coordinator Salem Commitment Tracking Coordinator

LR-N14-0146 Attachment #1 Attachment #1 Description and Assessment

LR*N14-0146 Attachment #1 LAR S14-02 License Amendment Request to Revise Technical Specifications (TS) to Adopt TSTF-51 0, Revision 2, "Revision to Steam Generator Program I nspection Frequencies and Tube Sample Selection" Table of Contents 1.0 DESCR IPTION................................................................................................................ 2

2.0 ASSESSMENT

................................................................................................................ 2

3.0 REGULATORY ANALYSIS

............................................................................................. 3 3.1 No Significant Hazards Consideration Determination.............................................. 3 4.0 ENVIRONMENTAL EVALUATION.................................................................................. 5 1

LR*N14*0146 Attachment #1

1.0 DESCRIPTION

LAR S14-02 The proposed change revises the Technical Specification (TS) req uirements for Salem Generating Station Units 1 ahd 2, regarding steam generator tube inspections as described in TSTF-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection." The proposed change also removes unnecessary information related to the original Salem Unit 2 steam generators which were replaced in 2008. The proposed changes are necessary to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications.

The proposed amendment is consistent with TSTF-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation PSEG has reviewed TSTF-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," and the model safety evaluation as part of the Federal Register Notice of Availability (76 FR 6 6763). PSEG has concluded that the justifications presented in TSTF-51 0 and the model safety evaluation prepared by the NRC staff are applicable to Salem, Units 1 and 2, and justify this amendment for the incorporation of the changes to the Salem Units 1 and 2 TS.

The traveler and model safety evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). Salem was not licensed to the 10 CFR 50, Appendix A, GDC. The Salem equivalent of the referenced GDC 14, 15, 30, 31, and 32 are the AEC Proposed General Design Criteria (July 1967) Criteria 9, 33, 34, 35, 36 as described in Section 3.1 of the Salem Updated Final Safety Analysis Report (UFSAR). This difference does not alter the conclusion that the proposed change is applicable to Salem Units 1 and 2.

2.2 Optional Changes and Variations PSEG is not proposing any variations or deviations from the TS changes described in the TSTF-51 0, Revision 2, or the applicable parts of the NRC staff's model safety evaluation.

Salem, Units 1 and 2 TS utilize a different numbering system than the Standard Technical Specifications (STS) NUREG-1431 on which TSTF-510 was based.

Specifically, the "Steam Generator (SG) Program" in the Salem Unit 1 TS is numbered 6.8.4.i rather than 5.5.9, the "Steam Generator Tube Integrity" TS is numbered 3.4.5 rather than 3.4.20, and the "Steam Generator Tube I nspection Report" is numbered 6.9.1.1 0 rather than 5.5.9. The "Steam Generator (SG) Program" in the Salem Unit 2 TS is numbered 6.8.4.i rather than 5.5.9, the "Steam Generator Tube Integrity" TS is numbered 3.4.6 rather than 3.4.20, and the "Steam Generator Tube Inspection Report" is numbered 6.9.1.1 0 rather than 5.5.9. These differences are administrative and do not affect the applicability of TSTF-51 0 to Salem Units 1 and 2.

2

LR-N14-0146 Attachment #1 LAR S14-02 In addition, PSEG requests to revise the Salem Generating Station Unit 2 Technical Specifications (TS) 6.8.4.i- "Steam Generator (SG) Program," 6.9. 1.10- "Steam Generator Tube Inspection Report," and the Bases Section of 3/4.4.6 "Steam Generator (SG) Tube Integrity," to remove unnecessary information related to theW* distance that is only applicable to the original Salem Unit 2 Westinghouse steam generators. PSEG replaced the Westinghouse steam generators with Areva steam generators in 2008.

TS 6.8.4.i "Steam Generator (SG) Program", proposed changes revise the following:

1. TS 6.8.4. i.c (page 6-19d) removes "The following alternate tube criteria... " down to 6.8.4.i.d.

2. TS 6.8.4.i.d (page 6-19e) deletes sentence "The portion of the tube within the hot-leg tubesheet region below theW* distance is excluded."

3. TS 6.8.4.i.d (page 6-19e) deletes "Note: Step 2 has two separate requirements (a and b), depending on the type of SG tubes installed."

4.

TS 6.8.4. i.d.2a (page 6-19e) deletes item 2a.

5. TS 6.8.4.i.d.2b (page 6-19e) revises the numbering format from 2 b to 2.

6. TS 6.8.4. i.d.4 (page 6-19f) deletes item 4 TS 6.9. 1.10 "Steam Generator Tube Inspection Report", proposed change revises the following:

1. TS 6.9.1.10.h (page 6-24b) deletes item h TS Bases Section 3/4.4.6 "Steam Generator (SG) Tube Integ rity", proposed changes revise the following:

1. Bases Section 3/4.4.6 (page B 3/4 4-3) deletes sentences "The portion of the tube w ithin the hot-leg tubesheet region below theW* distance is excluded. The excluded portion of the tube defined by W* is ONLY applicable to Westinghouse Model 51 SGs with mill annealed Alloy 600 tubing expanded into the tubesheet using the Westinghouse explosive tube expansion (WEXTEX) process."

2. Bases Section 3/4.4.6 (page B 3/4 4-3c) delete the paragraph starting with "License Change Request (LCR)... S05-07" through the last paragraph on page B 3/4 4-3e.

Pages B 3/4 4-3d and B 3/4 4-3e will be removed from the TS Bases.

The Bases markup pages are being p rovided for information only.

3.0 REGULATORY ANALYSIS

---311'--Jo Sig11ificat:1t 1:-lazar:ds Cot"lsidemtiof"l Salem, Units 1 and 2 requests adoption of an approved change to the standard technical specifications (TS) into the plant specific technical specifications (TS), to revise the Administrative Controls, 6.8.4.i, "Steam Generator (SG) Program," and Specification 6.9. 1.1 0, "Steam Generator Tube Inspection Report," and LCO 3.4.5 (Salem Unit 1) I 3.4.6 (Salem Unit 2), "Steam Generator Tube Integrity," to address inspection periods and other administrative changes and clarifications.

As required by 10 C FR 50.91 (a), an analysis of the issue of no significant hazards consideration is presented below:

3

LR-N14-0146 Attachment #1 LAR 514-02

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 p robability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to the Salem Unit 2 Technical Specifications (TS) that are not associated with TSTF-51 0, removing unnecessary information related toW* that is only applica ble to Westinghouse steam generators, is an administrative change that does not involve a s ignificant 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 Steam Generator 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.

The proposed changes to the Salem Unit 2 Technical Specifications (TS) that are not associated with TSTF-51 0, removing unnecessary information related toW* that is only applicable to Westinghouse steam generators, is an administrative change that does not affect the design of the SGs or their method of operation.

--+J:Jet"efGI"e,it-is-cGRGiuded-tl:lat-tnese-cbar:ges-do-noLcr:eateJbe_possibility_otaJJew_or-different kind of accident from any accident previously evaluated.

3.

Do the proposed changes 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 pressu re 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 4

LR-N14-0146 Attachment #1 LAR 814-02 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.

The proposed changes to the Salem Unit 2 Technical Specifications (TS) that are not associated with TSTF-51 0, removing unnecessary information related to W* that is only applicable to Westinghouse steam generators, is an administrative change that does not involve a significant reduction in a margin of safety.

Therefore, it is concluded that the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the above, PSEG 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.

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 C FR 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 C FR 51.22(c)(9). Therefore, pursuant to 10 C FR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

5

LR-N14-0146 Attachment #2 Attachment #2 Mark-up of Proposed Technical Specification Pages

LR*N14*0146 Attachment #2 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License DPR-70 are affected by this change request:

Technical Specification 3.4.5, Steam Generator (SG) Tube Integrity 6.8.4.i, Steam Generator (SG) Program 6.9. 10, Steam Generator Tube Inspection Report 3/4 4-7 6-19b 6-19c 6-19d 6-24a The following Technical Specifications pages for Renewed Facility Operating License DPR-75 are affected by this change request:

Technical Specification 3.4.6, Steam Generator (SG) Tube Integrity 6.8.4.i, Steam Generator (SG) Program 6.9. 10, Steam Generator Tube Inspection Report 3/4 4-9 6-19b 6-19c 6-19d 6-19e 6-19f 6-24a 6-24b

!yEACTOR COOLANT. SYSTEM STEAM GENERATOR j_SG)

TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 SG tube integrity shall be maintained and all SG tubes s<;tipf.ying the tube repair criteria shal l pe plugged in accordance with the.Steam Gener*tor Program.

APPLICABILITY:

MODES 1, 2,

3 and 4.

ACTION:

a. *

.

With one or more SG tubes..ps.atisfying the tube :repair criteria and not plugged in accordance with the Steam.Gen.erator Program:

1.

Verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection within 7 days; and

2.

Plug the affected tube (s) in accordan(Je with t):le.Stt3C<ffi Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or. SG tube inspection.

b.

With SG tube integrity not maintained or the,xequired Action of a.

above not met I be in at least HOT STru:JPBX _wi t:ii'in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least COLD SHUTDOWN within the follo.wing3o hours.

SURVEILLANCE REQUIREMENTS

4. 4. 5.1 Verify SG tube integrity in accor.dance with.. il.e Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies criteria is plugged in accordance.. with the Steam Generator entering HOT SHUTDOWN following a SG tl.,\\be inspection, plugging

Separate Ac tion is allowed for eachSG tube.

SALEM -

UNIT 1 3/4 4-7 Amendment No. 268

DMINISTRATIVE CONTROLS

7) Limitati ons on the dose rate resulting fr9ll1. fA¢t:i,qqtive mf.?,l;t?rtal released in gaseous effluents to areai( beyond* the SITE BOUNpA:RY Coriforining to the doses associated wi th 10 CFR *Part 20, Appen:dix B,.Table II.,

Column 1,

8) Limitations on the annual and quarbedy air doses resulting froin noble gases released in gaseous effluents from each ui:iit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part $._0,.

9) Limitations on the annual and,, quarterly dqses *to a MEMBER OF THE PUBLIC from Iodine:..:131, Iodine-133 I tritiUm,.arid all'radionuclides iri 'pa:itlcuiate form wi th halfli ves greater than 8 days in gasiiious effluents releiifi!ea from each unit to areas beyond the SlTEBOUNDARY conforming to Appe'ridi:X: I to 10 CFR Part 50,

10)

Limitations*on the annual dose or dose..... cotru:nit'ini'mt*to.anyMEMBER OF THE PUBLIC due to releases of. radioactivity and to

radiation from urariimn fuel cycle sources conforming to 40 CF.R Part 190.

6. 8. 4.h Radiological Environmerital Monitoring Ptc:>9:t"aJ,U A program shall be provided to monitor the radii:i:ti.on and radionuclides in the enviro.ns of the plant.

The prOgram shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2). verification of the accuracy of the "(i£Hueiit morii\\:.orfri.g program and modeling of environmental e:x:posurepat'hways; The.program::;Jhall (1) be contained in the ODCM, (2) conform to the guidanCe of Ai;:>peiidi:X:I to 10 CFR Part so, and (3) include. the following:

1)

Monitor.ing,

sampling, analysis, and reporting of radiat ion and radionuclides.in the environment in accordanc.e with the methodology and parametersin the ODCM,

2)

A Land Use Census to ensure that-ohanges in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications \\:.o the monitoring progn\\m are IJ1ade if. required by the results of the census, and

3)

Participation* in a Interlaboratory Comparison P.rogramto erJ.sure that independent checks on the preqision and adcu:Citcy oftbl measurei:ii.ents of radioactive materials.. in environmEmtal sample matrices are perf'ormed as part of the quality assurance. program for envirbnmental' monitoting.

6.8.4.i Steam Generator (SG)

Program A Ste am Generator Program shall be established and. iinpleriiei:ib:id to ensure

wa csGtube:r-in t egriT:y-i s maint"."a-tnea. -.. J;n a:a.ai:t:ion.l *. tnest:eam *Grie-rato*"':i:---

Program sht;l,ll include the following 'W.*i:1

a.

Provisions for condition monitoring *a.ssessmmts.

Condition monitoring assessmentmeans an evaluation ofthe*"as found" condition of the tubing with respect t.o the performance cri teria for str.uctural integrity and accident induced leakage.

The "as found".condition refers to the condition of.th:eHtubing during an SG inspection outage;: as determined f:t:Om'the ihservice inspection results or by otl).er means, prior to. the plugging of tubes.

Condition moni to;r-ing MSessmeQ.tR. s.hall be conduc.t.ed during each SALEM UNIT 1 6-19b

. Amendment No. 268

ADMINISTRATIVE CONTROLS outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

Performance criteria for SG tube integrity. SG tube integrity shall be maintained by

"""A-:cll:-:i-n _ s _e-rv--= i-c _e _s7 te_ a_ m___, meeting the performance criteria for tube structural integrity, accident induced leakage, generator tubes and operational leakage.

b.

shall retain

1.

Structural integrity performance criterion: All in serviee steam generator tubes structural integrity sA-all retain struclliral integrity over the full rangWJ:7erating eondftioos over the full range ofll(including startup, operation in the pov.'er range, hot standby, and cool down and normal o perating all Xnticipat?d.trt;msients inek;!d?d in the design specification_) and design basis conditions (including ace1dents. fh1s mcludes retam1ng a afety factor of 3.0 agamst burst under normal steady state full power operation pnmary-to-secondary pressure startup, operation In differential and a safety factor of 1.4 against burst applied to the design basis the power range, hot accident primary-to-secondary pressure differentials. Apart from the above standby, and cool requirements, additional loading conditions associated with the design basis d ow n), all accidents, or combination of accidents in accordance with the design and anticipated licensing basis, shall also be evaluated to determine if the associated loads t

. nt i

1 d d contribute significantly to burst or collapse. In the assessment of tube integrity, ransle s

. nc u e those loads that do significantly affect burst or collapse shall be determined and In the des1gn assessed in combination with the loads due to pressure with a safety factor of 1.2 s pecification, and on the combined primary loads and 1.0 on axial secondary loads.

design basis accidents.

2.

Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 1 gallon per minute per SG.

3.

The operational leakage performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System 0 erational Leakage."

plugging

c.

Provisions for SG tube Fepaif critena.

u es found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate F9f*NF criteria shall be applied as an alternative to the 40% depth based criteria:

1.

Tubes with service-induced flaws located greater than 15.21 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located-in-the-portion-oHhe-tube-from-the*top-oHhe tubesheet-to-15÷21-inehes---*

below the top of the tubesheet shall be plugged upon detection.

d.

Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube

criteria.

SALEM - UNIT 1 Amendment No. 303

ADMINISTRATIVE CONTROLS The portion of the tube below 15.21 inches from the top of the tubesheet is excluded from this requirement.

The tube-to-tubeshoet wold is not part of the tube. In addition to meeting tho 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. Afl assesment of degradation shall be performed to determine the type and location of I' s to which th ubes may be susceptible and, based on this assessment, to determi e which inspection ethods need to be employed and at what locations.

degradation assessment

1.

2.

INSERT: "affected and potentially affected"

3.

results in m ore frequent inspections Inspect 100% of the tubes in each SG during the first refueling outage following SG replaoemeAt:0installation. I Inspect 100% ofthe tubes at sequential periods of 120, 90, and thereafter, 60 effective full power montt:ls. The first sequential period st:lall be considered to egin after-the-fi.Fst inserviee inspeet+oo-e1/23/4ct 50% of the tubes by the refueling outage nearest the midpoint of the period and the FeFAaffitr[by-tfltH:efue\\oo]*FC&-the-e-nG-ef-tAe-per-ioEh-NB-8G shall operate for more than 48 effective full power months or t\\vo refueling AHH'l'SH'HWHI.Gfw..tef-If crack indication found in portions of the SG tub not excluded above, then the next inspection for ea SG for the degradation m chan ism that caused the crack indication shall not exceed 24 effective full pow r months or one refueling outage (whichever *

). If definitive information, s ch as from examination of a pulle

!agnostic non-destructive testing, or e gineering evaluation 1cates that a crack-like indication is not associated ith a crack(s), then the indication need not be treated as a crack.

e.

Provisions for monitoring operational primary-to-secondary le kage.

DELETE AND REPLACE FROM INSERT #1 SALEM

UNIT 1 6-19d Amendment No. 303

I nsert #1 Salem Unit 1 6.8.4.i.d.2 I page 6*19d "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 that 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 inspected period may be prorated. The fraction of locations to be inspected for this potentia l 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

--_,p-erin-ds-:"

ADMINISTRATIVE CONTROLS

2. WCAP-8385, Power Distribution Control and Load Following Procedures - Topic al Report,

( Propriet ary)

Methodology for Specification 3/4.2.1 Axial Flux Difference.

3. WCAP-10054-P-A, westinghous e Small Break ECCS Evaluation Model Using NOTRUMP Code (W Proprietary),

Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor.

4. WCAP-10266-P**A, The 1981 Version of Westi nghouse Evaluation Model Using BASH Code, (W Proprietary)

Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor.

5. WCAJ?-12472-P-A, BEACON - Core Monitorins and Ope rations Support

System,

( Proprietary).

6. CENPD-397-J?-A, I mp rov ed Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Tec hnology.

c.

The core operating limits shall be determined such that all applicable limits (e.g.,

fuel thermal mechanical li mits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS)

limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycl.e to the NRC.

6.9.1.10 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry i nto HOT SHUTDOWN following completion of an inspection performed in accord anc e with the Specification 6.B.4.i, "Steam Generator (SG)

Program."

The report shall include:

DELETE

a.

The scope/spections performed on each SG,

.gr-tion mechanisms found, 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

, active d egradati on mechanism,

f.

tubes plugged to date, and

g.

The r es ult s of con 'tion monitoring, including the results of tube pulls and in-s itu t ti ng.

SALEM -

UNIT 1 The n u mber and percentage of tubes pl ugged to date, and the effective plugging percentage in each steam generator, and 6-24a Amendment No. 284

REACTOR COOLANT SYS'l'EM 3/4.4.6 STEAM GENERATOR (SG)

TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.6 SG tube integrity shall be maintained and all SG tubes satisfying the tube be plugged in accordance with the Stearn Generator APPLICABILITY:

MODES 1,

2, 3

and 4.

ACTION:

a.*

With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Stearn Generator Program:

1.

Verify tube integrity of the affected tube(s) is maintained until the next re fueling outage or SG tube inspection within 7 days; and

2.

Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.

b.

With SG tube integrity not maintained or the required Action of a.

above not met, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.1 4.4.6.2 Verify SG tube integrity in accordance with the Steam Generator Program.

Verify that each inspected SG tube that satisfies the tube criteria is plugged in accordance with the Stearn Generator prior to entering HOT SHUTDOWN following a SG tube inspectio Separate Action is allowed for each SG tube.

SALEM UNIT 2 3/4 4-9 Amendment No.

262

ADMINISTRATIVE CONTROLS

7)

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1, B)

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

9)

Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium,

and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Pa.rt 50,

10)

Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

6.8.4.h Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.

'l'he p rogram shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accura cy of the effluent monitoring program and modeling of environmental exposure pathways.

The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1)

Monitoring,

sampling, analysis,

and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2)

A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of the

census, and

3)

Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6. 8.4.i 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 Generat.or Program shall include the following r&V*ien:

SALEM UNIT 2 6-19b Amenrunent No.

262

ADMINISTRATIVE CONTROLS

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 plugg1ng 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 a:r:*e being met.

b.

All in service steam generator tubes shall retain structural Performance criteria for SG tube integrity.

SG tube integri.ty shall be maintained by meeting the perfo.t*mance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1.

Structu.tal integrity p erfo:r.mance criterion:

A+/--+/--in service integrity OVer the full s-t-e-am-generator tubes sha-H-r-e-t;.a.;L--ft--54:SE:1*t:ural inte-!;j-F"'o'e-E range Of normal ffie---:ftd:-1 range of ROX:ffit-l-e-j7CEating conditions (incluel-ifi§'

operating conditions _j fl-; epe-ration in the f*lW<a-r-.,-t-aT}

a.:nd.*-eoe-:1 (in cluding startup down and all anticipated transients included in the design specificat+/-ett7 and design basis accidents.

This includes operation In the power retaining a s afety factor of 3. 0 against burst under normal range, hot standby, steady state full power operation primary-to-secondary pressure and cool down), all differential and a safety factor of 1. 4 against burst applied a nticipated transients t the deign basis accident primary-to-scondary presuc dÿfferentals.

Apart from the above reqturements, addV tlonal Included In the des1gn loading conditions associated with the design basis accidents, s pecification, and or combination of accidents :i.n accordance with the design and design basis licens ing basis, shall also be evaluated t o determine if the accidents 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.

c.

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 t.he leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 1 gallon per minute per SG.

3.

The LCO specified :i.n Leakage."

Tubes found by inservice equal to or exceeding 40%

be plugged.

SALEM - UNIT 2 6-19c Amendment No.

262

ADMINI STRATIVE CONTROLS

!DELETE I d.

1. WEXTEX expanded region inspection metho dology (W* Methodol ogy ).

This alternate repair criteria is only appli cable to Wes tinghous e Model 51 S Gs wi th mill anne a l ed Al loy 600 tubing expanded into the tube sheet using the Wes tinghouse explos ive tube expansion ( WEXTEX) p roces s.

The de finitions that apply to W* are provided b e l ow :

so*ttom o f WEXTEX trans ition

( BWT )

i s the hi ghe s t point o f cont act b etween the tube and t h e tube s heet a t,

or below the top-o f-tubeshee t,

as determi ned by eddy current tes ting.

W* Length is de fined a s the length o f tubing b e l ow the bottom of the WEXTEX trans i tion

( BWT) that mus t be demonst :r:a ted to be non-degraded in o rder far the tube to maintain s tructural and leakage integri t y.

For: the hot leg, the W* l ength is 7. 0 inche s,

which repre s ents the mos t cons ervat ive hot l e g l ength.

W* Dis tance is de fined as the non-degraded di stance from the top of the tubesheet to the bottom o f the W*

l ength,

including the dis tance from the top-o f-tubesheet to the bottom of the WEXTEX t rans ition

( BWT )

and Non-Dest ructive Examination ( NDE )

mea s urement un certainties

( i. e.,

W*

dis tance "' W* l ength + di s tance to BWT + NDE uncertainti es ).

The W* Di s t ance is t h e la rger of the fol lowing two dis t ances as meas ured from the top-of-the tub e s heet

( TTS ) :

( a )

8 -inches below t he 'l'T S or

( b )

the non degraded di s tance from the TTS to the bottom of the W*

l ength,

incl udin g the dis tance from the TTS to the bottom o f the WEXTEX trans ition

( BWT )

and Non-Dest ructive Examination (NDE) measurement uncertainties

( i. e.,

W*

distance

= W*

length + di s tance to BWT + NDE uncertaint i e s )

Tubes within the hot-leg region o f the tubesheet wi th flaws identified in the W* Di stance, shall be removed from s ervice on detection by t ube pl uggin g.

Flaws l o cated belovl the W*----

di s tance wi thin the hot-leg regi on o f the tubesheet may remain in s e rvi ce regardl e s s of s i z e.

Provi s i ons for SG tube inspect i ons.

Period i c SG tube inspe cti ons shall be performed.

The number and portions of the tubes inspected and methods o f inspection shall be p e rformed with the obj e ctive of detecting flaws of any type ( e. g.,

volurnet:cic flaws,

axial and ci r cumfe rential cracks )

that may be present along the SALEM UNIT 2 6 - 1 9 d Amendment No.

2 6 2

ADMIN I STRATIVE CONTROLS l e ngth of the tube,

f rom the tube-t -tub e s heet weld i n l e t t o the tube-to-tubesheet wel t the tube outl may s ati s fy the app l i cab l e tube + criteri a.

!DELETE 1--1--) the tube with:in the hot leg tubeshect regie i-s-e-:zte+/-ooecl.

The tub e - t o-tubesheet wel d is not part degradation assessment iDELETE j 1--

DELETE AND REPLACE FROM INSERT #2 In addition to meeting the requiremen t s o f d. l, d. 2,

below, the inspection s cope,

inspection methods,

and inspection intervals shall be such as to ens ure that SG tube integrity i s maintained un*til the next S G inspect i on.

assessment of degradation shall be performed to determin he type and location of flaws to which the tubes may be s u s cepti le and, based on this a s s e s sment,

to det ermine whi ch inspect ion m. thods need to be emp loyed and at what l o cati ons

DELETE "n" 1.

Inspect 100 % o f the tube s in e a ch S G during the first fol lowing SG two s eparate requi rements the type o f SG tubes ins talled.

2 a.

SGs with Alloy 6 0 0 Mill Anneal ed tub es :

Inspect 1 0 0 % o f the tubes a t s e quential periods o f 60 e ffect ive full power month s.

'rho first s equential period s hall b e considered to begin after the first inservice inspe cti on o f the SGs.

No SG shall operate fo r mo re than 24 effective full powe r months o r one refueling outage

( whi chever is les s ) without being inspected.

scs Hith Alloy 690 Thermally Treated t\\::l:hes-!

Inspect 100%

t:he tubes at sequential periods of lH1 :J:-0-8-; 9-7!:-; e:ttd thereafter,

e.G effecti'lf-full pm<cr months.

The first 1---_.,.':;,.. sequential poriod---erfta-.1-1 be ceFI£-i.Ele-red to bcgii'l afte.r the first

i::J.'l;&e-E'.l-i-ee in sp e et i e-l'l:

a f the--B*EJ.s.

f.n--addi t i on,

kn-speet-§-0-'lr---e-f o tubac by the rcfHelinutagc nerest the midpoint of the period and the remaining 50% ey the refueling outage lrea:t'est 8-1!d of the pcrioE!.

Ne-te for morc--th-a effective full poHer menths or three refueling outages (Hhichcver is less) -,dthout being inspcct:cd.

3.

I f crack indications a rc found in any S G tub e, then the next inspection for each SG for the deg radation me chanism that caused the crack in ica tion sha l l not exceed 2 4 effective full power months or one refueling outage

( whichever

).

I f INSERT: "affected and potentially affected" de fini tive informat on,

such as from examinati on o a pulled tube,

di agnostic non-de s t ructive t e s t ing, or engi ecring evaluation indi cates that a crack-like indi cation is not a s s ociated with a crack ( s ), then the indication n.ed not be treated as a

crack.

SALEM -

UN IT 2 results in more frequent i nspections Amendment No. 2 62

I nsert #2 Salem Unit 2 6.8.4.i.d.2 / page 6*19e "After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third 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, c and d 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 144 effective full power months. This constitutes the first inspection period; b)

During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c)

During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d)

During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent insp-ection penods."

4.

When the W* methodology has been implemented, Inspect 1 00 percent of the lnservlce tubes for the entire hotleg tubesheet W*

distance with the objective of detecting flaws that may satisfy the applicable tube repair criteria of TS 6. 8.4. 1.c. 1 every 24 effective full power months or one refueling outage (whichever is less).

e.

Provisions for monitoring operational primary-to-secondary leakage.

6. 8.4.j lnservice Testing Program 6.8.4.k This Program provides controls for In service testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a.

Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASM E OM Code and applicable Addenda terminology for inservlce testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years Required Frequencies for performing inservice testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 1 84 days At least once per 276 days At least once per 366 days At least once per 731 days

b.

The provisions of Specification 4.0.2 are applicable to the above required frequencies and to other normal and accelerated frequencies specified as 2 years or less In the lnservice Testing Program for performing inservice testing activities,

c.

The provisions of Specification 4.0.3 are applicable to inservice testing activities, and

d.

Nothing in the ASME OM Code shall be construed to supersede the requirements of any Technical Specification.

Reactor Coolant Pump Flywheel Inspection Program

---. --I l n-additlen-te-the-reqldirements-of-tl:1e-ISI-Rrogram,-each-Reactor_coolantt>ump_ -1/2-flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1. 1 4, Revision 1, August 1 975. I n lieu of Position C.4.b(1) and C.4.b(2), a qualified inplace UT examination over the volume from the inner bore of the flywheel to the circle one"half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels m ay be conducted at 20 year intervals.

SALEM UNIT 2 6-1 9f Amendment No. 281

ADMINIS'l'RATIVill CONTROLS 2.

WCAP - 8 3 8 5, Power Distribut i on Control and Load Foll owing Procedures Topical Report,

(!_Proprietary )

Methodology for Specif ication 3 I 4. 2. J.

Axial Flux D i f f erenc e 3.

WCAP - 1 0 0 5 4 - P - A, West inghous e Small Break illCCS Evaluat ion Model Usins NOT RUMP Codt,

{

Proprietary ),

Methodology for Specif ication 3 / 4. 2. 2 Heat Flux Hot Channel Fac tor.

4.

WCAP - 1 02 6 6 - P -A, The 1 9 8 1 Version of Westinghouse Evaluation Model Using BASH Code,

(! Proprietary ) Methodology for Specification 3 /4. 2. 2 Heat Flux Hot Channel Factor.

5.

WCAP 1 2 4 7 2 - P -A, BEACON Core Monitoring and Operation Support gysteJn, ( Proprietary).

6.

CENPD - 3 9 7 *** 1? - A,

ImEroved Flow Measurement Accuracx using Crossflow Ultrasonic Flow Measurement Technol ogy 7.

WCAI? - 1 0 0 5 4 - 1? - A, Addendum 2,

"Addendum to the wes tinghou s e Small Break ECCS Evaluation Model Using the NOTRUMP Code :

Saf ety nj ection into the Broken Loop and COS I Condensat ion Model. "

c.

The c or e operat ing l imi t s sha l l be determined such that all app li cable limi t s

{ e. g., fuel thermal mechanical limits, core thermal hydraulic limits,

Emergency Core Cool ing Systems (ECCS )

limit s, nuclear limits such as SDM 1 transient analysis limits,

and acc ident analys i s limits ) of the saf ety analys i s are met.

d.

The COLR,

including any mid** cycle revis ions o :r: supplements shall b e provided upon i s suance f or each reload cycle t o the NRC.

6. 9. 1. 1 0 STEAM GENERATOR TUSE INS PECTION REPORT A report shall be submitted within 1 8 0 days after the initial entry into HOT SHUTDOWN following comp letion of an inspect ion performed in accordance with the Specification 6. 8. 4. 1, steam Generator

( SG)

Program. u The report shall include :

a. The scop e of inspections performed on each SG,

b.Yi""' ion mechani s'"" found,

+ sondestruct ive examinat ion techniques u t i l ized for each degradation nteohani am, d. loocation,

orientation

( i f l inear),

and mea sured s i zes (if available) of service induced indications,

=1/4 SALEM UNI'l' 2 6 - 24a mendment No.

2 6 7

ADMINISTRATIVE CONTROLS The number and percentage of tubes plugged to d ate, and the effective plugging percentage in each steam generator, e.

. * *Numbe:t of tubes plugged

degradation mechanism,

the inspect ion outage for each aebive

f.

Total number and percentage of tubes plugged to date,

g.

monitoring,

including the results of tube pulls regarding the appl ication of W* inspection methodology

. ( applicable to tubes within the hot - leg region of the tube sheet ) ;

including the number of indications,

the location of indications

( :relative to the BWT and TTS ),

the orientation (axial,

circumferent ial, volumetri c ),

the severity of each indicat ion

( e. g., near through-wall or not through wall ),

the tube side where the indication initiated ( inside or outside diameter),

the cumulative number of indications detected in the tubesneet region as a function of elevation within the tubesheet,

the condition monitoring and operational assessment main steam line leak rate ( including aggregate calculated mai.n steam line break leak rate from all other sources ), and an assessment of whether the :results were consiStent with expectations

regarding the number of flaws and flaw s everity {and if not consistent,

a des c:dption of the proposed corrective action).

6.. 2 Special reports shall be submitted to the u. s. Nuclear Regulatory Colllt!liSsion,

Document Control Pesk,

Washington, P. C. 2 055 5, with a copy to the Administrator, USNRC Region I within the time period specified for each report.

6 * $J ; 3 PELETEP 6. Sl. 4 When a report is required by ACTION 8 OR 9 of Table 3. 3 11 '1Accident Monitoring Instrumentation ", a report shall be submitted within the following 14 days.

The :report shall outline the p:replanned alternate method of monitoring for inadequat e care cooling,

the cause of the inoperability,

and the plans and aoh$dule for restoring the instrument channels to OPERABLE s tatus.

SALEM - UNlT *2 6 - 24b Amendment No.

275

LR-N14-0146 Attachment #3 3/4------

Attachment #3 Mark-up of Proposed Technical Specification Bases Pages

LR-N14-0146 Attachment #3 Mark-up of Proposed Technical Specification Bases Pages The following Technical Specifications Bases pages for Renewed Facility Operating License D PR-70 are affected by this change req uest Technical Specification Bases B 3/4.4.6, Steam Generator (SG) Tube Integrity B 3/4 4-2 B 3/4 4-3 B 3/4 4-4 B 3/4 4-4a The following Technical Specifications Bases pages for Renewed Facility Operating License DPR-75 are affected by this change request Technical Specification Bases B 3/4.4.6, Steam Generator (SG) Tube Integrity B 3/4 4-3 B 3/4 4-3b B 3/4 4-3c B 3/4 4-3d B 3/4 4-3e

REACTOR COOLANT SYSTEM BASE S 3 / 4. 4. 4 PRES SURI ZER The l imit on the maximum wat e r volume in the pre s suri z e r a s sures that the parame t e r is ma i n t ained within the normal s t e ady-state enve l ope o f operation a s s umed i n the SAR.

The l imit i s con s i s t ent with the i n i t i a l SAR as sumpt ions.

The Surve i l l ance Frequency i s b a s e d on operating experienc e,

e quipment r e l i abi l i t y,

and pl ant r i s k and is cont r o l l e d under the Surve i l l ance Frequency Control Pr ogram.

The maximum wat e r volume a l s o ensures that a s t e am bubbl e i s formed and thus the RCS i s not a hydraul i c a l l y s o l i d s ys t em.

The requ i rement that a minimum numbe r of pre s sur i z e r h e a t e r s be OPERABLE a s sure s that the pl ant w i l l be able to e s t ab l i s h natural c i r cu l a t i on.

3 / 4. 4. 5 S TEAM GENERATOR

( S G )

TUBE INTEGRITY The LCO requires that SG tube integ r ity be ma int ained.

The LCO a l s o r e qu i r e s t h a t a l l SG tub e s t h a t s a t i s f y t h e reBtir cri t e r i a be plugged in a.;;;,

cco rdance with the S t e am Generator Pro gram.

I I 1

plugging P uggmg Dur ing an SG i

pection,

any inspected tube that s a t i s f i e s the S t e am Generat o r Program cri t e r i a i s removed from s e rvi ce by plugging.

I f a tube wa s determi ned to s a t i s f y the c r i t e r i a but was not plugged, the tube may s t i l l have t ube integr i t y.

I n the context o f this Spe ci f i c a t i o n,

a SG tube i s defined a s the entire l ength o f the t ub e,

including the t ub e wall between the tub e - t o -tub e sheet we l d a t the t u b e i n l e t a n d t h e tube-t o-t ub e s h e e t weld at t h e tube outlet.

Tubes with s e rvice-i nduced fl aws l ocat e d greater than 1 5. 2 1 i nche s b e l ow the top of the tub e s h e e t do not require pluggi ng.

Tub e s with s e rvi c e - i nduce d f l aws l ocated i n the portion o f the tube from the t op of the tub e s h e e t t o 1 5. 2 1 inch e s b e l ow the t op of the tube sheet s h a l l be plugged upon de t e c t i on.

The tub e - t o - tube s h e e t weld i s not con s i de r e d part o f the tub e.

A SG tube has tube integrity when i t s a t i s fi e s the SG per formance c r i t e ri a.

The S G per formance c r i t e r i a are defined in Speci fication 6. 8. 4. i,

" S t e am Gener a t o r

( SG )

Program, "

and describe acc ept able S G tube performanc e.

The S t e am Generator Program a l s o provi d e s the evaluat i on proce s s for d e t e rmi ning conformance with the SG p e r fo rmance c r i t e r i a.

There are three SG per formance c r i t e r i a :

s t ructural integri t y,

accident i nduced l e akage,

and ope r a t i onal l e a kage.

Fai lure to me e t any one o f these criteria is con s i de red failure t o me e t the LCO.

The s t ructural integrity p e r formance c r i t e ri on provide s a ma rgin of safety against tube bur s t o r c o l l ap s e unde r norma l and accident condi t ions,

and ensures s t ructural i n t e g r i t y o f the SG tub e s under all anticipated transie nt s i ncluded in the de s ign s p e c i f i ca t i on.

Tube bur s t is defined as,

" The gross s t ructural fai lure of the tube wa l l.

The condi t i on typical l y corre sponds to an unstable opening d i sp l a c ement

( e. g., opening area incre a s e d i n r e sponse t o constant

-- -----*- ¸ Å p

""r-e s sure )

accompanied by duc t i l e

( p l a s t i c )

t e aring o f the tube mat e r i a l at the ends o f the degradat ion. "

Tube col l ap s e is defined a s,

" For the l oad di spl ac ement curve for a given s t ructur e,

col l ap s e occurs a t the t op o f the l oad versus di spl acement curve where the s l op e of the curve b e come s z e ro. "

The s t ructural i n t e g r i t y performance c r i t e r i on provi des guidance on as s e s s ing loads that s i gni f i cant l y affect bur s t or c o l l ap s e.

I n that context,

the t e rm

" s i gn i f i cant" i s defined a s,

"An accident l oading cond i t i on other than d i f f e rent i al p r e s sure is con s i de r e d s i gn i f i cant when the addi t i on of such l oads i n the a s s e s sment of the s t ructural i n t e g r i t y per formance c r ite rion could cau s e a

l ower s t ructura l l imi t or l imi t i ng bur s t / c o l l apse c ondi t i on t o b e e s t ab l i shed. "

SALEM -

UN I T 1 B

3 / 4 4 - 2 Amendment No.

3 0 3

( P SEG I s sued )

BASES 3 /4. 4. 5 STEAM GENERATOR

( SG ) TUBE INTEGRITY

( Cont inued)

The determination of whether thermal loads are primary or sec ondary J.oads is based on the ASME definit ion in whi ch secondary loads are s e l f - l imi t i ng and wi l l not cause fai lure under s ingle load app l i c a t ion.

For tube integrity evaluations,

except for c i rcumferential degradat ion, axial thermal loads are c l a s s i f i ed as s e c ondary loads.

For c i rcumf erential degradat i on,

the clas s i f i cation of axial thermal loads as primary or secondary loads wi l l be evaluated on a case -by-case bas i s.

The divi s i on be tween primary and secondary c l a s s i f i cat ions w i l l be bas e d on detail ed analy s i s and/ or t est ing.

Structural integrity requi res that the primary membrane s tress intens ity in a tube no t exceed the yi eld strength for all ASME Code,

Section I I I,

Service Leve l A

(normal operat ing condi t ions )

and.Service Level B (up s et or abnormal condi t i ons )

transients included in the des ign spec i f i cation.

Thi s inc ludes s afety factors and app l i cable d es ign bas i s loads based on ASME Code,

Sect i on I I I,

Subsection NB and draf t Reg.

Guide 1. 12 1.

The accident induced l e akage performance cri t eri on ensures that the primary to secondary l eakage caused by a des ign basis acc ident,

other than a s te am generator tube rupture

( SGTR ),

i s wi thin the accident analysis assumptions.

The accident analy s i s as sume s that accident induced leakage doe s not exceed 1 gpm per SG.

The accident induced l eakage rate inc ludes any primary - to - s econdary leakage exi s t ing prior to the accident in addJ.tion to primary - t o - s ec ondary l eakage induced during the acc ident.

The operat ional l e akage performance c riterion provides an observab le indi cat i on of 8G tube condi t ions during p l ant op e ration.

The l imit on operat i onal leakage is contained in LCO 3. 4. 6. 2,

"Op erat ional Leakage,

and l imits primary - to secondary l e akage through any one B G t o 15 0 gallons p e r day.

Thi s limit is bas ed on the as sumption that a s ingl e c rack le aking thi s amount would not propagate to a

SGTR under the stress condit ions of a LOCA or a main s te am l i ne break.

I f thi s amount of l eakage is due to more than one crack,

the cracks are very smal l, and the above a s sumpti-on is cons ervative.

'I'he ACTION requi r ement s are modified by a Note clari fying that the Ac t ions may be entered independently for each SG tube.

This is acceptable becau s e the ACTION requirements provide appropriate compensatory actions for each af fected SG tube.

Complying with the ACTION requirements may al l ow for continued operat ion,

and subsequen.

G tube s are governed by subs equent ACTION requiremen t s.

plugging I f it i.s dis covered that tube s examined in an inserv:L c e inspect i on sati sfy the tube cri teria bt we re not p lugged i² accordance with the Steam Generator Program, an evaluat ion of SG tube integrity of the affected tube ( s ) must be made.

Stearn generator tube integrity i s based on me et ing

- - ¹- º * -- >> - c:n:e-sG-pe:D:'formEmcre--cr it*eT ta-d*e*s*cr lhed-+/-n*the-st*eam-Genera tor-Program-;-The-s&-S criteria define l imit s on SG tube degradat ion that allow for flaw growth een inspect i-ons whi l e s t i l l p roviding assurance that the SG performance 1.!---"=--='-' cri teri a wi l l cont inue to be met.

In order to determine i f a SG tube that should have been plugg ed has tube integrity, an evaluat i on must be comp l eted that dernonst:rates that the SG p erformance criteria wi ll cont inue to be met unt il the next refue l i ng outage or SG tube inspect ion.

The tube integrity determination i s bas ed on the e s timated condi t ion of the tube a t the t ime t he s ituat ion i s d i s c overed and the est imated growth of the degradat ion prior t o the next S G tube inspec tion.

An action t ime of 7 days is sufficient to compl et e the evaluat ion whi l e minimiz ing the :dsk of pl ant operat J.on with a SALEM -

UNIT 1 B 3 / 4 4 - 3 Amendment No.

2 7 9

( PSEG i ss ued)

plugging REACTOR COOLANT SYSTEM BASES 3 / 4;. 4. 5 STEAM GENERATOR

( SG) TUBE INTEGRITY

( Continued)

SG tube that may not have t11be integr i ty.

I f the eval.uation determines..that the af fected tube ( s ) have tube integrity, plant operati'on i s allowed to continue unt i l the next refuel ing outage or SG inspection provided the.

inspection interval continues to be supported by

_ an operat ional as sessment that reflects the affected tubes.

However,

the affected tube ( s ) ll\\1'!:\\t.. l:?.e plugged prior to entering HOT SHUTDOWN fol lowing the next refue l ing outage or SG inspection.

'rhi s act ion t ime i s accep tabl e since operation unt i l.. the. next inspection is supported by the operational assessment ;

If SG tube integri ty is not being maintained or the ACTION. requirements are not met, the reactor must be brought to HOT STANDB:Y: wi.tbin.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD.SHUTDOWN wi thin 3 6 ho1,1rs.

The <;.ction time s are reasonable based on operating experienc e,

to reach the desired plant cond:i. tiona.* f*ra*m iuu power conditio.ns.in an orderly. manner and without challengihg plant syst;.ems.

During shutdown periods the SGs are inspected as. required. by su:r:vei l l anc= requ i rements and the steam Generator Program.

NEI

97 - 0 6, "Steam Generator Program Guideline s, 11 and its referenced EPRI Guidel ines 1. es tab_l sh the content of the Steam Generator Program.

Use of the Steam Generator Program ensures that the inspection is appropriate and consistent.1Nith accepted indus try practices.

During SG i nspections a condition monitoring assessment of the SG tubes, is performed.

'rhe condi t ion moni taring as ses sment determines.tl:le

".as found" condit ion of the SG tube s.

'rhe purpose of the condi tie>:n monitoring asse13 sment is to ensure that the.. 8G performance criteria have been met. for t.he previous operating pe riod.

The Steam Generator Program determin!s. _the scope of the nspe on an ethods used to determine whether. the tiibes con tain flaws satisfying the tube Criter ia Inspect ion scope (L.. e.. 1. whioh t:uJ::>.es or areas of tubing wi thin the SG are to be inspecte¢1.)

is a funct ion of f;lxis ting and pot ential degradation locations.

The Steam Gene.ra.to.r:,.J?rogram als.o. :

spec ifies the inspect ion methods to be used to f ind eii sting and potential degradatipn.

Inspe c tion methods. are a function of deg}'adaHon, mor.ph!:l:.lo_gy I nondestructive examinat ion

( NDF.)

technique capabil ities at1d inspection locat ions.

The Frequency i s determined by the operat ional asses srnent.and other l imi ts iii the SG. examination guide l ines The Steam Generatoo:r:. J?rogram u s e s information on exi s t ing degradat ions and growth rates to dete.rrriine an inspection Frequency that provide s reasonable as ;;;ur,CI:Q_QFJ _ t;,J:lat: tb,e. tu,b.ing wil l..

meet the SG performance cri teria at the next sche(:'lulgd inspect ion '.

:Cn addi t ion,

specif:icat ion.6. 8. ). i contains prescript;i:ve: e(JJ.lirements concerning inspect ion intervals to provide added as surance tha,t

..tl:l.e SG perf.o:).::mance l-----.,cr". :LceriawnJ-:--J:remecnet.we

- -eli--:J.n-:-spece:i:o.:o&,,,..

plugging ct. ion,

any J.nSp<:=cted tube thci.'

saU s f i e s

the Steam Gene r r Program criteria is removed from servi e by plugging.

The tube repair cri.t:eria.delineated in Spe d f:L cat iqn.. !L*;s :.ji:;;i

te inten.4,ed to ensure that tubes accepted for continued service sat i t;>Y

'he SG performance cri teria. with allowance for error in s i z e. measu:( INSERT: If crack indications are fou nd in

addition, the tube repair c ri teria,

in conj uncti s t eam G(O)nerator Progr 1

ensure that the SG perf any SG tube, the maximum inspection to be me.t un.ti.l the n xt inspection o f the subj e interval for all affected and potentially provide s guidance for perfor.ming operational a s s tubes remaining i n s e vi ce wi l l cont inue to meet affected SGs is restricted by Specification The Frequency of prio

.. ent e ring HOT SHUTDOWN 6.8.4.i u ntil subsequent inspections support extending the inspection interval.

SALEM UN I 'l' 1 *

B 3./4.4 - 4 TSBC SCN 0 5 04 8

REACTOR COOLANT SYSTEM BASES

/ 4. 4. 5 S'rEAM GENERATOR

( SG)

TUBE INTEGRITY

( ContinueQ.)

¶ e*ure s that tll.E:l.surve i l l apqe has been comp leted and a l l tubes meet ing the repair criteria are. plugged prior to sub j e c t ing the SG tubes to s i gnificant primary-to - sec.onciary pres sure different ial.

3 / 4. 4. 6 REACTOR, COOLANT SYSTEM LEAKAGE 3 / 4.. 4.. 6.. 1.

L:EAKl\\(}E DETECTION SYSTEMS.

The RCS l eakage detect ion systems required by this specif icatibn f3.re provided to monitor and detect leakage f rom the Reactor coolant J?rflS$U.:t;:e Boundary.

The se detect ion systems are cons i s t ent with the reco:mmemdations of Regulatory Guide 1. 4 5,

" Reactor Coolant Pres sure Boundary Leakage Detect ion systems ", May 197 3.

3 / 4. 4. 6. 2

.OJ?E:RATIONAL LEAKAGE Industry experience has shown that whi le a l :i.mited amount of leah;age i s expected from the. RCS,

the unidenti f ied port i on o f thi s leakage can b e. reduced to a thre shold value.of less than 1 GJ?M.

This threshold valqe ¨ s ' stifficiently low to ensure. early detection of additional leakage.

The 1 0 GPM IDENTIFIED LEAKAGE limitation provides allowance : :Eor. a. l imited amount of leakage from known sources whose presence will not inter:E¢te wi th the dete ct ion of UN!DENTIFIE:D LEAKAGE by the l eakage detection sys tems.

PRESSURE BOUNDARY LEAKAGE of any magnitude i s unacceptable.s in.ce i t may be indic ative of an impending gross fail ure of the pres sure boundary.

Therefore,

the presence of any PRESSURE BOUNDARY LEAKAGE requi res i;:,he

t),ni t to be promp t ly p laced in COLD SHUTDOWN.

Pd:..l!l§I:EY'" to-Secondary Leakage Thr.ough Any one SG The primary-to-secondary leakage rate limi t applies to leakage through any one Steam Generator ;

The limi t of 1 5 0 gal lons per day per. s team *gerti3i-&:tor

i s * **

based. on the operat ional l eakage performance criterion in: NEI 9 7 - 0 6., St;e.S,m Generator Program Guidel ine.s.

The Steam Generator Progra.!XL..QpeJ:;:§.tJ9nal leakage performance cri terion in. NEI. 97 0 6 state s,

"The RCS operational primary-to secondary leakage through any one SG. sha l l be limi ted to 1 5 0: gallons per day. u :

The. l imit is. pased on operating experience with steam geni:e;(,:J.tor tube degra¢lat ion mecha,ni f3tns that resu l t in tube leakage.

. Th<:J gp_era.tional leakage rate criterion. jn; conj unct ion w i th the implementation.. of the Steam Generator

--dp=:cogra_m_i-;s---an--ffeut-tvedmeasuredfor-minimizing-the--frequ,en.gy;:C?ft::::;team e --c---

generator tube ruptures.

Act ions Unidenti fied leakage or

ident i f ied Ieakage in excess of. thi :LCO limi t s mus t be reduced to.wi thin l imits within 4 hou.:r.s.

This action tim allows.time to ver i fy leakage rates and _ e it her identify tmident i f iecl Je'){a,ge*... 61;' reduce leakage to within l imi t s.before the reacto r mus t be shut down.

' This act;:;i,on j, s neces sary to prevent further deterioration of the reactor.:cool ant pressur.e boundary (RCPB ).

I"f any pr_es sure bounda ry leakage exis ts,

or pr imary - t o secondary leakage is not. within l i mi t, o r i f unidenti fied d.:t'.i.denti fied.

Ieakage

cannot -be reduced t o wit hin J,imits wi thin '* hours, he rea.ctor must be brought to lower pres sure c*onditions to reduce the severity 6f the lea)i;:().ge and SALEM -

UNIT 1 B 3 / 4 4 - 4 a TSBC S CN 0 5.- 0.4 8

REACTOR COOLANT SYSTEM BASES 3 / 4. 4. 5 RELIEF VALVE S ( continued)

B.

Automatic cont rol o f PORVs t o control reactor cool ant s ys t em pres sure. This is a function that reduces challenges to the code safety valves for overpressurization events,

including an inadvertent actuation of the S a fety Injection System.

c.

Maintaining the integrity o f the reactor cool ant pressure boundary.

This is a

function that is related to controlling identi fied leakage and ensuring the ability to detect unidentified reactor coolant pressure bounda ry l eakage.

D.

Manual control o f the block valve to

( 1 ) unblock an i s olated PORV to allow it t o be used for manual and automati c control o f Reactor Coolant System pres s ure

( I tems A & B ),

and ( 2 ) is ol ate a

PORV wi th exce s s i ve s e a t leakage

( I tem C ).

E.

Manual control o f a block valve to i s ol ate a s tuck-open PORV.

3/4. 4. 6 STEAM GENERATOR

( SG)

TUBE INTEGRITY There are three SG performance criteria :

s tructural integrity, accident induced leakage,

and ope rational leakage.

Failure to mee t any one of the s e criteria is cons idered failure to meet the LCO.

SALEM UNIT 2 B 3/4 4-3 Amendment No. 262

( PSEG I s sued)

REACTOR COOLANT BAS ES I f it is di s cove re more SG tubes examined in an i s ervice inspectio n s ati s fy the tube criteria but we re not plugged in a ccor ance wi th the S team Generator P rogram, an evaluation of SG tube i ntegrity o f the f fected tube ( s ) mus t b e made.

Steam generator tube integrity i s b a s ed on meeting e SG pe rformance c riteria de s cribed in the Steam Gene rator P rogram.

The SG criteria define limits on SG tub e that allow for flaw growth between inspections whil e s till p roviding a s s urance that the SG performance criteria wi l l continue to be met.

I n o rde r to determine i f a SG tube that s hould have been plugged ha s tube integrity, an e valuation mus t be comp l eted that demons trates that the SG pe rformance criteria wi ll continue to be met until the next re fueling outage o r SG tube inspection.

The tube i n tegrity dete rmination is based on the estimated condi ti on o f the tube at the time the si tua ti on is di s covered and the estimat ed growth of the degradati on p ri or to the nex t SG tube inspection.

An a c t i on time o f 7 days i s s ufficient to complete the evaluation whil e minimi zing the ris k of plant op eration wi th a SG tube that may not have tube integrity.

I f the evaluation det.ermi11es that the a f fe cted tube (.s )

have tube integrity, plant op eration i s all owed to continue until the next re fueling outage or SG inspecti on provided the inspection interval continues to be supported by an operational a s s e s sment that re flects the a f fected tubes.

Howeve r, the a f fe cted tube ( s ) mus t be plugged prior to ente ring HOT SHUTDOWN foll owing the next refueling out age or SG inspection.

Thi s allowed outage time is a cceptable s ince opera tion until the next inspection i s supported by the operational as ses sment.

I f SG tube inte g rity i s not being maint ained or the Action requi rements are not met, the reactor mus t be brought to HOT STANDBY withi n 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN wi thin 3 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The act ion times a re reas onable b a s ed on operating exp e rience, t o reach the des i red plant conditions from full powe r condi tions in an orde rl y manner and without chall enging plant sys tems.

During s hutdown peri ods the SGs are inspe cted as requi red by s urveillance requi rements and the Steam Generator P rogram.

NEI 9 7 - 0 6,

"St eam Generator Prog ram Guidel ines, a and i t s refe renced EPRI Guidelines, es tablish the content o f the Steam Generator P rogram.

Use of the Steam Generator Program ensures that the inspection i s app r opriate and cons is tent with accepted industry practices.

plugging During SG inspections a condi tion monitoring asses sment o f the SG tube s i s performed.

The condition monito ring asses sment determines the "as found" of the SG tubes.

The purpose of the condi t i on moni to ring as ses sment is t

that the SG performan ce cri teria have b een met for the p revious ope rating The Steam Generator Program determines the s cope o f the inspection and t used to determine whether the tub e s contain flaws s atis fying the tube k

-*---* -* ---cr

-it;e,;.ia.µ -I-nspection-s cope

-(.i-, e-.-,-wb.i.Gh-tubes-or-a :r;eas -o-f-tubi-ng -with-i-n-the- -SG-al:'e-- -

to be inspected) is a function of existing and potential degradat i on l ocations.

The Steam Generat or Program also speci fi es the inspecti on methods to be us ed to find potential degradation.

Inspection methods are a function of degradation morphol o gy, nondes truct ive examination ( NDE) technique capabilities and inspection locations.

The Frequency is dete rmined by the operational as s e s sment and other l imi ts in the SG examination guidelines.

SALEM -

UNIT 2 Amendment No.

2 6 2

( PSEG I s sued)

REACTOR COOLANT S Y STEM INSERT : " If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification BAS ES 6.8.4.i u ntil subsequent inspections support

======="===-=='1==9 extending the inspection interval.

T h e S t eam Generat or Program us es informat i o r a t e s to de termine a n inspection Fre quency the tub i ng will me et the SG pe r fo rmance cri In additi on,

Spe c i f i cation 6. 8. 4. 1 contains inspection interva ls t o p rovide added a s s u wi l l be met between s cheduled inspections.

exi s ting degradat i ons and growth p rovides reas onable a s surance that eria at the next s chedul ed inspe ct io n.

p res c riptive requir ements concerning n ee that the SG performan ce criterria-plugging nspec Lon, any i n spected tube that s at i s fies the Steam P rog1*am,..,.4-;- c ri teria i s removed from s e rvi ce by plugging.

c r i t e r i a delineated in Speci fica tion 6. 8. 4. i a re intended to ensure that t ubes a ccepted for continued s e rvi ce s a ti s fy the SG performance criteria wi t h a l lo nee for e rror i n s i ze meas urement and fi.1ture gr owth.

In addition, the tube c r i t e ri a,

in conj unction wi th other element s of the S t eam Generator Program, ensure t hat the S G per formance cri t e ri a wi l l continue to be met unt i l the next inspection o f the subj ect tub e ( s ).

NEI 9 7 - 0 6 provide s guidance for performing operational a s s e s sments to

,he tubes rema i ning in s e rvi ce wi l l cont inue to meet the SG p e rfo rmance c

plugging e

Frequency of p rior to ent ering HOT SHUTDOWN following a

SG inspectio ensures that the Surveillance has b een completed and al l tubes mee ting the criteria a re plugged prior to s ubj ecting the SG t ubes to s i gn i f i cant rimary-to - s e conda ry pre s s ure di fferential.

Li cen s e Change Reque s t ( LCR) S O S-07

( LR-N O S - 0 3 97, LR-N 0 6 - 0 2 77, LR-N 0 6- 03 3 8 )

provides requirements for limited tubesheet insp e ction that i s only appli cable withi n the hot leg WEXTEX expanded region o f the tubesheet for the Sa lem Uni t 2 Wes t inghous e Se ri es 5 1 S team Generators.

LCR S 0 5- 0 7 i s s uppo rted by, but not limi ted t o, the guidance provided in WCAP-1 4 7 9 7, Revi s i on 2, "Generi c W* Tube Plugg ing Cri teria for 51 Series Steam Generator Tub e s heet Re gion WEXTEX Expans i ons "

and suppo rt ing in formation p rovided from We s tinghous e Lett e r Report LTR-CDME-05-30,

"W*

Int egri ty Evaluation for S al em Unit 2 Limited SG Tube RPC Examination

( B a s ed WCAP-1 4 7 9 7, Revis ion 2 ).

In acco rdance with LCR SOS- 0 7, the W* Length i s the undegraded l ength o f tubing into the tube s he et be low t he bo ttom o f the WEXTEX t ran s i t i on

( BWT )

that preclude s tube pull out in the event o f a complete circumfe rential s eparation o f the tube below the W* Length.

For the hot l e g, the W* Length is 7. 0 inches,

whi ch repres ents the mos t conservative hot leg l ength defined in WCAP-1 4 7 97, Revi sion 2.

The W*

D i s tance is the l a rger of the f o llowing two di s tances as measured from the t op-o f-the-tub e s heet

( TT S ) :

( a )

8 - inches below the TTS or

( b )

the non-degraded dis t ance from *the TTS to the bottom o f the W*

Length,

including the distance from the TTS to the bottom of the WEXTEX trans i ti on

( BWT )

and Non-Des t ructive Examinat i on (NDE )

measurement uncertaint i e s

( i. e., W

Di s tan ce "" W"' Length + dis tance t o BWT + N D E uncertaintie s ).

Non-De s t ructive Examinat ion determines the di s t ance to the BWT for each tube.

The nonde s t ructive

-exami-nat-1-e n-(*NDE)--measu.r;emESnt-unce-J;tain.ty-i s-p rovided-from-.LCR _ S.05=0_7_,_a s_s.upp_o_r:t_e.d _

by WCAP - 1 4 7 97 Revi s ion 2.

Tube s wi th indi cations detected within the W*

Di s t ance wi ll be removed from s e rvi ce by tube p lugging.

SALEM -

UNI T 2 B 3 / 4 4 - 3 c Amendment No.

2 6 2

( P S EG I s s ued)

REACTOR COOLANT SYSTEM BASES Tube degradat i on of any type o r extent below the W* Di s tan ce,

including a complete c i r cumferential s ep arati on o f the tub e, i s a cceptable and there fore may.temain in s e rvi ce.

As applied a t Salem Unit 2, LCR S 0 5 - 0 7 i s used to de fine the re qui red tube in specti on depth into the tube s heet, and i s not u s ed to pe rmit degradation in the W* Di s tance to remain in s e rvice.

Furthe rmore,

potenti al p rima ry to s econdary l eakage in the W* Dis tance,

and b e l ow the W* Di s tan ce,

can be cons e rvatively evaluated in accordance with LCR S O S - 0 7.

The leak rate potential for axi a l,

ci.tcumferenti a l, and volume t r i c indi cations detected wi thin 12 inches from the top of the tubesheet can be cons e rvatively calculated using the cons trained crack model a s delineated in LCR S O S-07 ( supported by Westinghous e LTR-CDME-0 5 - 3 0 ).

The pos tul a t ed leakage during a s t eam line b reak shall be equal to the foll owi ng

equation, as supported by LCR 3 0 5 - 0 7 :

P ostulated SLB Leakage

=

As s umed Lea kage

>12"

<TTs As sumed Leakage o"-s" <TTs + As sumed Le a kage B"-lZ" <TTS +

Where :

Ass umed Leakage o"-8" <TTS i s the postulated l eakage for indi cations that a re deemed via f l aw depth e s t imat i on techniques to be 1 0 0 % throughwal l,

and the refo re p re s ent a potential leak path.

This term i s applicabl e t o detected indi cations during a n in-servi ce inspection and potentia lly undetected indi cations in the s team generator t ubes l e ft in s e rvi ce between 0 inches and 8 inches b elow the t op of the tube sheet

( TT S ).

Since tubes with indi cations detected betwe en 0 and 8 inches below the TTS a re plugged upon detection, the calculation of thi s t e rm for the ass e s sment of SLB leakage for the subs equent operation cycl e foll owing an in-s e rvi ce inspe ction only requi res consideration of potentially undetected indi cati ons.

The cal cul ation o f thi s t e rm for the a s s e s sment of SLB l e akage fo r the p revi ous ope rat ion cycle,

f o l lowing an in-s ervi ce inspection, requires considera tion of both detected and potentiall y undetected indi cations.

As sumed Leakage "B-12" <TTS i s the conse rvatively proj ected leakage in

steam-'3ene :r;-a-tortubls4between-8-and-412-inGhes-be-1ow--t-hl4t-op-of-the-----

tub e s heet.

Implementation o f LCR S 0 5 0 7 does not requ i re tube i nspection below the W* Distance,

there fo re the methodology for conservatively cal culating the population o f indications between 8 and 12 inches bel ow the TTS is provided by fitting a SALEM -

UNIT 2 B 3 / 4 4 - 3d

.....,0-E-LE_T_E_P_A_G_E__,J Amendment No. 2 62

( P SEG I s s ued)

REACTOR COOLANT SYSTEM BASES regre s s ion line to the cumulative inspection data

( dete cted indi cations )

from all SGs and pro j ecting the numbe r o f indi cations ( to minus 1 2 inches below TTS )

us ing a 9 5-percent probability predi ction bound.

The cumul ative indi cations from all s t e am generators are conservatively as s umed to occur in one SG

( s imi lar to figure 1 6 of Westinghous e LTR-CDME- 0 5 - 3 0 ),

The conserva tive leakage rate for the indicati ons b etwe en 8 and 12 inches is 0. 0 03 3 gpm multiplied by the numbe r of p roj e cted indications

( a s dis cu s s ed in LCR S 0 5 - 07 s ubmitta l s LR-N06-0277 and LR-N0 6- 0 3 3 8 ).

The leak rate o f indi cations dete cted between 8 and 1 2 inches are bounded b y the proj ected total di s cu s s ed above,

a s s uming that the inspect i on resul t s fo r detected indi cations d o not contradi ct the calculated population a s des cribed p revious ly.

Assumed Leakage

>l2" <'ITS is the cal culated leakage from the s team generator tubes l e ft in servi ce below 12 i nches from the top of the tubeshee t.

Thi s is 0. 0 0 0 0 9 gpm times number of tubes left in servi ce in the s t eam generator.

Each SG i s as sessed for Main Steam Line Break

{MSLB )

l eakage individual l y in accordance wi th the di s cuss ion above, and the SG with the mos t calculated leakage i s cons ervatively assigned as the a ffe cted S G.

The cal culated MSLB leakage provided above,

including MSLB le akage from a l l other s ources,

shall be reported t o the NRC in accordance with applicable Techni cal Speci fi cations.

The Calculated MSLB Leakage must b e less than the maximum allowabl e MSLB leak rate limit in any one s team generator in orde r to maintain doses within 10 CFR 5 0. 67 guideline values and within GDC-19 values during a postulated main s team line break event.

SALEM UNIT 2

!DELETE PAGE B 3 / 4 4-3e Amendment No.

2 62

( PSEG I s s ued)

v t e Letter Sequence Request
TAC:MF5117, (Approved, Closed)
Initiation Request, Request Acceptance... Supplement Administration Questions 24 March 2015 Answers 21 May 2015 Results Approval Other:
MONTHYEAR2015-03-24 [Table View]

ML14210A484

Text

Subject:

2.0 ASSESSMENT

3.0 REGULATORY ANALYSIS

1.0 DESCRIPTION

2.0 ASSESSMENT

3.0 REGULATORY ANALYSIS

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