Text

License Amendment Request Nos. 324 and 196 Re Steam Generator Tube Integrity
	ML053140195
Person / Time
Site:	Beaver Valley
Issue date:	11/07/2005
From:	Lash J; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-05-144
	Download: ML053140195 (102)
	v • d • e

FENOC FENOCRoute Beaver Valley Power Station 168 RPO. Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077.0004 James 11. Lash 724-682-5234 Site Vice President Fax: 724-643-8069 November 7, 2005 L-05-144 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 License Amendment Request Nos. 324 and 196 Steam Generator Tube Integrity Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) requests amendments to the above licenses for Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2 in the form of changes to the Technical Specifications. These License Amendment Requests (LARs) are being made to satisfy a commitment to submit a LAR to adopt steam generator technical specification requirements consistent with Technical Specification Task Force 449, "Steam Generator Tube Integrity." The commitment was made in a letter dated April 13, 2005 (L-05-069) that submitted Beaver Valley Power Station Unit No. I LAR 320, "Replacement Steam Generators" for NRC review. It is expected that LAR 320 will be implemented in the spring of 2006, prior to approval of LAR 324 in the summer of 2006.

The proposed license amendments revise the requirements in Technical Specifications related to steam generator tube integrity. The changes are consistent with NRC approved Technical Specification Task Force Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4. The availability of this technical specification improvement was announced in the Federal Register on May 6, 2005, as part of the consolidated line item improvement process (CLIIP).

The FENOC evaluation of the proposed changes is presented in the Enclosure. The proposed Technical Specification changes are presented in Attachments A-1 and A-2 of the Enclosure. Attachment B-1 and B-2 of the Enclosure provide the proposed information-only changes to the Technical Specification Bases related to the proposed license amendments.

AOD\

Beaver Valley Power Station, Unit Nos. 1 and 2 License Amendment Request Nos. 324 and 196 L-05-144 Page 2 The BVPS review committees have reviewed the changes. The changes were determined to be safe and do not involve a significant hazard consideration as defined in 10 CFR 50.92 based on the safety evaluation and no significant hazard evaluation described in the Enclosure.

FENOC requests approval of the proposed amendment by July 1, 2006, with a 60 day implementation period to precede the Fall 2006 Beaver Valley Power Station Unit No. 2 refueling outage scheduled for October 2006.

No new commitments are contained in this submittal. If you have questions or require additional information, please contact Mr. Gregory A. Dunn, Manager, Fleet Licensing at 330-315-7243.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November 7 , 2005.

Sincerely, me.Lash

Enclosure:

FENOC Evaluation of the Proposed Changes.

c: Mr. T. G. Colburn, NRR Senior Project Manager Mr. P. C. Cataldo, NRC Senior Resident Inspector Mr. S. J. Collins, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

ENCLOSURE FENOC Evaluation of the Proposed Changes Beaver Valley Power Station License Amendment Requests 324 (Unit 1) and 196 (Unit 2)

Subject:

TSTF-449, "Steam Generator Tube Integrity" Table of Contents Section Title Page

1.0 INTRODUCTION

....................................... 1

2.0 DESCRIPTION

OF PROPOSED AMENDMENT ............................. 1

3.0 BACKGROUND

....................................... 2 4.0 REGULATORY REQUIREMENTS AND GUIDANCE ................... 2

5.0 TECHNICAL ANALYSIS

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

6.0 REGULATORY ANALYSIS

............. ........................... 2 6.1 Verification and Commitments ......................................... 3 7.0 NO SIGNIFICANT HAZARDS CONSIDERATION ......................... 6 8.0 ENVIRONMENTAL EVALUATION ........................................ 6 9.0 PRECEDENT ........................................ 6

10.0 REFERENCES

.. 7 Attachments Number Title A-1 Proposed Unit 1 Technical Specification Changes A-2 Proposed Unit 2 Technical Specification Changes B-I Proposed Unit I Technical Specification Bases Changes B-2 Proposed Unit 2 Technical Specification Bases Changes C Comparison to TSTF-449 i

Beaver Valley Power Station License Amendment Requests 324 (Unit 1) and 196 (Unit 2)

1.0 INTRODUCTION

This License Amendment Request (LAR) is being made to satisfy a commitment to submit a LAR to adopt steam generator technical specification requirements consistent with Technical Specification Task Force (TSTF) 449, "Steam Generator Tube Integrity." The commitment was made in a letter dated April 13, 2005 (L-05-069) that submitted Beaver Valley Power Station Unit No. 1 LAR 320, "Replacement Steam Generators" for NRC review. It is expected that LAR 320 will be implemented in the spring of 2006, prior to approval of LAR 324 in the summer of 2006.

The proposed license amendment revises the requirements in Technical Specification (TS) related to steam generator tube integrity. The changes are consistent with NRC approved Technical Specification Task Force Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4. The availability of this technical specification improvement was announced in the Federal Register on May 6, 2005, as part of the consolidated line item improvement process (CLIIP).

2.0 DESCRIPTION

OF PROPOSED AMENDMENT Consistent with the NRC-approved Revision 4 of TSTF-449, the proposed TS changes include:

Revised TS definition of LEAKAGE

Revised TS 3/4.4.6, "Reactor Coolant System Leakage"

New TS 3/4.4.5, "Steam Generator (SG) Tube Integrity" replacing old TS 3/4.4.5 "Steam Generators"

New TS 6.19, "Steam Generator (SG) Program"

New TS 6.9.7, "Steam Generator Tube Inspection Report" Proposed revisions to the TS Bases are also included in this application. As discussed in the NRC's model safety evaluation, adoption of the revised TS Bases associated with TSTF-449, Revision 4 is an integral part of implementing this TS improvement. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

Page 1

Beaver Valley Power Station License Amendment Requests 324 (Unit 1) and 196 (Unit 2)

Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2 Technical Specifications are currently based on NUREG 0452 rather than NUREG 1431 which is the format of TSTF-449. Therefore, Attachment C has been provided to correlate changes proposed by the TSTF with those proposed for BVPS by this license amendment request.

To meet format requirements the Index, Technical Specifications and Bases pages will be revised and repaginated as necessary to reflect the changes being proposed by this LAR.

3.0 BACKGROUND

The background for this application is adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

4.0 REGULATORY REQUIREMENTS AND GUIDANCE The applicable regulatory requirements and guidance associated with this application are adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

5.0 TECHNICAL ANALYSIS

FirstEnergy Nuclear Operating Company (FENOC) has reviewed the safety evaluation (SE) published on March 2, 2005 (70 FR 10298) as part of the CLIIP Notice for Comment. This included the NRC staff's safety evaluation (SE), the supporting information provided to support TSTF-449, and the changes associated with Revision 4 to TSTF-449. FirstEnergy Nuclear Operating Company has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to BVPS Unit Nos. 1 and 2 and justify this amendment for the incorporation of the changes to BVPS Unit Nos. I and 2 TS.

6.0 REGULATORY ANALYSIS

A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability Page 2

Beaver Valley Power Station License Amendment Requests 324 (Unit 1) and 196 (Unit 2) published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

6.1 Verification and Commitments The following information is provided to support the NRC staffs review of this amendment application and reflects installation of replacement steam generators at BVPS Unit No. 1 planned for the spring of 2006:

Beaver Valley Power Station Unit No. 1 Steam Generator (SG) Model(s): Model 54F Effective Full Power Years (EFPY) 0.0 EFPY at start of Cycle 18 (Estimated start up date of service for currently installed SGs is 04117/06)

Tubing Material Inconel 690 Thermally Treated Number of tubes per SG 3592 tubes per SG Number and percentage of tubes SG 'A' - 0 tubes plugged (0.0%),

plugged in each SG SG 'B' - 0 tubes plugged (0.0%),

SG 'C' - 0 tubes plugged (0.0%)

Number of tubes repaired in each SG None. No tube repair techniques, other than tube plugging, are currently approved for these SG's Degradation mechanism(s) identified To date, no degradation mechanisms have been observed.

Current primary-to-secondary per SG: 150 gpd through any 1 SG leakage limits: Total: 1 gpm unidentified/l0 gpm identified Leakage is evaluated at: Room Temperature Approved Alternate Tube Repair Criteria (ARC):

There are no alternate tube repair criteria currently approved for these SG's.

Approved SG Tube Repair Methods:

There are no tube repair methods, other than tube plugging, currently approved for these SG's.

Page 3

Beaver Valley Power Station License Amendment Requests 324 (Unit 1) and 196 (Unit 2)

Beaver Valley Power Station Unit No. 1 Performance criteria for accident The primary to secondary SG tube leak rate value leakage assumed in the licensing basis accident analysis represents the operational value noted in LCO 3.4.6.2.c of 150 gpd per SG or a total of 450 gpd. The density used in the accident analysis to convert the volumetric leak rate to mass leak rate is 62.4 lbs/cu. ft.

Beaver Valley Power Station Unit No. 2 Steam Generator Model(s): Model 51M Effective Full Power Years (EFPY) 14.026 EFPY at end of Cycle 11 (04/04/05) of service for currently installed SGs Tubing Material Inconel 600 Mill annealed Number of tubes per SG 3376 tubes per SG Number and percentage of tubes SG 'A' - 151 tubes plugged (4.47%),

plugged in each SG SG 'B' - 160 tubes plugged (4.74%)

SG 'C' - 129 tubes plugged (3.82%)

Number of tubes repaired in each SG No tube repairs other than tube plugging have been made through Cycle 11 (2R1 1).

Degradation mechanisms identified

Tube wear at AVB intersections

Axial ODSCC at TSP intersections

Axial ODSCC in the hot leg sludge pile free span region and hot leg expansion transition

Axial PWSCC in the expansion transition and expanded portion in tubesheet

Axial PWSCC at dented tube support plate intersections

Circumferential ODSCC in the hot leg top of tubesheet expansion transitions

Axial ODSCC at freespan dings

Axial and circumferential PWSCC at the small radius U-bends

Volumetric OD indications above the top of tubeshect (non corrosion related)

Pitting above the TTS

Tube wear due to foreign object interaction Page 4

Beaver Valley Power Station License Amendment Requests 324 (Unit 1) and 196 (Unit 2)

Beaver Valley Power Station Unit No. 2 Current primary-to-secondary per SG: 150 gpd through any I SG leakage limits: Total: I gpm unidentified/10 gpm identified Leakage is evaluated at: Room Temperature Approved Alternate Tube Repair Criteria (ARC):

1. Voltage Based Repair Criteria Approved by:

Amendment #101, dated 08/18/99 Applicability (e.g., degradation mechanism, location):

ODSCC within Tube Support Plates Special limits on allowable accident leakage:

None identified Exceptions or clarifications to the structural performance criteria that apply to the ARC:

None identified Approved SG Tube Repair Methods Approved by:

1. Westinghouse laser welded sleeves 1. Amendment #56, dated 08/30/93

2. ABB/CE TIG welded sleeves 2. Amendment #98, dated 03/26/99 Applicability limits, if any:

1. None

2. None Sleeve repair criteria (e.g., 40% of the initial sleeve wall thickness):

1. 25% (Plug on detection of any degradation)

2. 32% (Plug on detection of any degradation)

Page 5

Beaver Valley Power Station License Amendment Requests 324 (Unit 1) and 196 (Unit 2)

Beaver Valley Power Station Unit No. 2 Performance criteria for accident The licensing basis main steam line break dose leakage consequence analysis has established a maximum allowable SG tube leakage of 2.8 gpm which includes an accident induced leakage of 2.5 gpm in the faulted SG. This total SG tube leakage value of 2.8 gpm includes the primary to secondary SG tube operational leakage noted in LCO 3.4.6.2.c of 150 gpd per SG.

Other accidents (except SG tube rupture) assume 150 gpd. The density used in the accident analysis to convert the volumetric leak rate to mass leak rate is 62.4 lbs/cu. ft.

7.0 NO SIGNIFICANT HAZARDS CONSIDERATION FirstEnergy Nuclear Operating Company has reviewed the proposed no significant hazards consideration determination published on March 2, 2005 (70 FR 10298) as part of the CLIIP. FirstEnergy Nuclear Operating Company has concluded that the proposed determination presented in the notice is applicable to BVPS Unit Nos. 1 and 2 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

8.0 ENVIRONMENTAL EVALUATION FirstEnergy Nuclear Operating Company has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298) as part of the CLIIP. FirstEnergy Nuclear Operating Company has concluded that the staff's findings presented in that evaluation are applicable to BVPS Unit Nos. 1 and 2 and the evaluation is hereby incorporated by reference for this application.

9.0 PRECEDENT This application is being made in accordance with the CLIIP. FirstEnergy Nuclear Operating Company is not proposing variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC staff's model SE published on March 2, 2005 (70 FR 10298). However, as stated in Section 2.0, DESCRIPTION OF PROPOSED AMENDMENT, Beaver Valley Power Station Unit Nos. I and 2 Technical Specifications are currently based on NUREG 0452 rather than Page 6

Beaver Valley Power Station License Amendment Requests 324 (Unit 1) and 196 (Unit 2)

NUREG 1431 which is the format of TSTF-449. Therefore, adaptation of TSTF-449 was required.

10.0 REFERENCES

Federal Register Notices:

Notice for Comment published on March 2, 2005 (70 CFR 10298)

Notice of Availability published on May 6, 2005 (70 FR 24126)

Page 7

Attachment A-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Changes License Amendment Request No. 324 The following is a list of the affected pages:

Page Pending LAR V

xv 1-3**

1-4 3/4 4-8 320 3/4 4-9 320 3/4 4-10 320 3/4 4-lOa 320 3/4 4-lOb 320 3/4 4-lOc 320 3/4 4-lOd 320 3/4 4-lOc 320 3/4 4-13 3/4 4-14 6-21 6-26 _

6-27*

6-28* _ _ _I

New page

- Provided for readability only

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4 .1 REACTOR COOLANT LOOPS 3/4.4 .1.1 Normal Operation .............. 3/4 4 -1 3/4 .4. 1.2 Hot Standby................. 3/4 4- 2b 3/4.4.1.3 Shutdown .................. 3/4 4-2c 3/4.4 .1.4. 1 Loop Isolation Valves - operating...... 3/4 4-3 3/4.4.1.5 Isolated Loop Startup............ 3/4 4 -4 3/4 .4.3 SAFETY VALVES................ 3/4 4 -6 3/4.4 .4 PRESSURIZER................. 3/4 4 -7 3/4 .4.5 STEAM GENERATORG (SG) Tube Integrity. 3/4 4 -8 I

3/4.4 .6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6. 1 Leakage Detection Instrumentation...... 3/4 4-11 3/4.4. 6.2 Operational Leakage ............. 3/4 4-13 3/4.4 .6.3 Pressure Isolation Valves.......... 3/4 4-14a 3/4.4. 8 SPECIFIC ACTIVITY.............. 3/4 4 -18 3/4.4. 9 PRESSURE/TEMPERATURE LIMITS 3/4.4 .9.1 Reactor Coolant System ........... 3/4 4-22 3/4.4 .9.3 Overpressure Protection Systems....... 3/4 4-27a 3/4.4. 11 RELIEF VALVES................ 3/4 4-29 BEAVER VALLEY - UNIT 1 VVAedetN. Amendment No. 24 2-44 1

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.8 PROCEDURES ........................................ 6-6 6.9 REPORTING REQUIREMENTS ............................ 6-17 6.9.1 DELETED 6.9.2 Annual Radiological Environmental Operating Report ......................... 6-17 6.9.3 Annual Radioactive Effluent Release Report ................................... 6-18 6.9.4 DELETED 6.9.5 Core Operating Limits Report (COLR) 6-18 6.9.6 Pressure and Temperature Limits Report (PTLR).6-20 6.9.7 Steam Generator Tube Inspection Report .6-22 6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM ..................... 6-21 6.12 HIGH RADIATION AREA .............................. 6-23 6.13 PROCESS CONTROL PROGRAM (PCP) .................... 6-24 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)........... 6-24 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS ................................ 6-25 6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM ......... 6-25 6.18 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM ......................................... 6-26 6.19 Steam Generator (SG) Program ..................... 6-27 BEAVER VALLEY - UNIT 1 XV Amendment No. 2 l

DEFINITIONS lreadabili ty only CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

LEAKAGE 1.14 LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be Pressure Boundary LEAKAGE, or BEAVER VALLEY - UNIT 1 1-3 Amendment No. 220

DEFINITIONS

3. Reactor Coolant System LEAKAGE through a steam generator to the secondary system (primary to secondary LEAKAGE).

b. Unidentified LEAKAGE Unidentified LEAKAGE shall be all LEAKAGE (except reactor coolant pump seal water injection or leakoff) that is not Identified LEAKAGE.

c. Pressure Boundary LEAKAGE Pressure Boundary LEAKAGE shall be LEAKAGE (except steam l generator tube primary to secondary LEAKAGE) through a nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

1.15 THROUGH 1.17 (DELETED)

QUADRANT POWER TILT RATIO (OPTR) 1.18 QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as

.the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The DOSE EQUIVALENT I-131 is calculated with the following equation:

CI 131DE =CI131 + C- 132 + CI- 1 3 3 + CI- 1 3 4 + CI- 1 3 5 170 6 1000 34 Where "C" is the concentration, in microcuries/gram of the iodine isotopes. This equation is based on dose conversion factors derived from ICRP-30.

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals; BEAVER VALLEY - UNIT 1 1-4 Amendment No. 2-14 1

l .. 1 - U l I i' .. 41 - ' I Y. 1'SM-Proposed changes to

) ,1A A . ST-SAM 1 BNFl1RAmPTtc draft page from Unit 1 LAR No. 320 LIMIITINGC CGDITION FOR oPERATION 3.1.5 Each steam generator shall be OPERABLE.

APPLICABILITY: P40DE£ 1, 2, 3 and 1.

With onc or more steam generators inoperable, rectore the inoperable generator(s) to GPERABLE status prior to increasing TM aboev 200 0 F.

SURVEILLANCE REQUIREMENT£ 4.4.5.1 Steam Cenerator- Gafnile- Seletien and Inpcin Eaeh steam generater shall be determined OPERABLE; durcing shutdewn by seleeting and inspecting at least the mninimfumf number of steam.

generaters specified in Table 4.1 1.-.

4.4.5.2 Steam Cenerater Tube Sample Selectien and Insi~eetien The tstam gnratr tubA minimum _ample sie, inAspAtiAn rsAult elassificatien, and the eorresponding aetien required shall be as epacified in Table 4.4 2. The inoe.iee inspeetien of oteam generater tubes shall be performed at the frequencies opecified in Gpeeifieatien 4.4.5.3 and the inspected tubes shall be varpCifiand aeceptable per the aeeptaneo eritaria of . peifi-at _ 1i T.4.5.4.

Steam generater tubes shall be examincd in aeeer-danee with Arctielea 0 of Seetien V ("Eddy current Thraminatien of Tubular- Preduets") and Appendix IV toe actien xi ("Eddy Cur rent Examinatien of-Nenferremagnetie Steam Canerater Ileat Se2~hanger Tubing") of the applicable year- and addenda of the AGME Boilar and Pressure Vessel Coda required by !GGFlRSO, Caction 50.55a(g) . The tubes salacted for eaeh inserv-1-e inspeetien shall include at least 3 percent of the total number of tubes in all steam generaterA the tubes 1elacted for thasa inspactiens shall be selected en a random basis emcapt.

a. Wherae ex-pariaenc in similar plants with similar- water-ehemiotry indicatas critieal areas to be inspacted, than at-leaast 50 parcant of the tubes inspacted shall be from these critical areas.

b. The firot sample of- tubes calacted foer eaah ins -----

inspactien (subsequent to the preservlee inspactien) of eaeh steam ganerater shall inelude-

1. Al! nonpiugged tubes that pr-evieusly had datactabla wall panetratiens greater than 20 parcant, and-

5:1#V1$'VALLL Ul'~l' 1

RBAGTOR COOLANT SYSTEM Proposed changes to jj draft page from Unit 1 LAR No. 320 SURVEILLANCE REQUIREMENTS (Continued) R o

2. Tubes in thosa areaa where emperianee has indieated_

petentiial preblems, and B. A tube insopetien pursuant to Speeifieatien I.4.S.1..a.8 shall be perfermed en eaeh zalected tube.

if any calected tube dees net permit the passage ef-the eddy current prebe for a tubeinspectien, thi shall be reerded and an adjacent tuba shall be zalaceted and oubjected to a tube inspeetien.

e. The tubes I I.

seleeted as the

_*- Iv seeond and Iaas Wo third samnples (if-re.uired By Table 4.. %2) during eaan inserviee insparzien may be subjected to a partial tube inspeetion provided:

1. The tubas salectId fer theas zsample include the tubzs frem these areaas of the tuba sheet array whera tubacs with imparfaetions were praevieuly found, and PhM-lh.. r nin' imparfationo on^ inc-lud wara ictlho-

_ravi-u-y

-or-timnc f ou nd of thc-' tiiumc^

where imper-ecetiens were pr-evieusly foeund.

P9P.179PT V?-TA.T-TP TATHTT I

n -7 e---- ---. T 7\ -_Tt .r~

fttjft= +/- UV: k= tl tj +/-:3y kIN t b It b +/- VIV Proposed changes to draft page from Unit 1 LAR No. 320 Thc rcoults of cach sample inspection shall bc classified into onc of thc following threc eategoric.

CatcrQorv Insicetion Re-ultq C 1: Less than 5 pe3rcent of the tetal tubes inspected arc degraded tubes and nonc of thc inspected tubcs arc defcetivc.

C 2 Onc or more tubes, but not morc than 1 pereent of thc total tubes inspected arc defcetivc, or between _ perecnt and 10 perecnt of thc total tubes inspected arc degraded tubce.

C 3 More than 10 pereent of the tetal tubes inospctca arc degradca eubce or morc tnan 1 pereent of thc inspected tubes arc defcetivc.

eotc; In all inspections, previously degraded tubes muot oehibit aignificant (grcatcr than 10 perecnt) further wall pcnctrations to bc included in the abovc pcrecntagc ealeulations.

4.41.5.3 inciueetien Free~uenelee The aboyc required inserviee inTpetiencs of steam generator tubcs shall be perfermed at the following frequencico.

a. The first inscrvice inspection of thc rlodel 54F stcam gencrators shall be performcd aftcr 6 Effcetivc Full Pewer Plonths but within 24 ealendar months of- initial criticality following stcam gencrator replacemcnt.

Subsequent inscrvicc inspections shall bc performcd at intcrvale of not less than 12 nor morc than 24 calendar months aftcr the previous inspection. If two conscutive inspections , not including thc prcscrvicc inspection, result in all inspeetien results falling into the C 1 eategery or if two eenseeutive inspeetiens demenstratce that previeusly ebserved degradation has not eontinuoed and no additional degradatien has oeeurred, the inspeetien intecval may be extended to a mamimum of enee per 410 monthso.

Nete. inservi e inspeetien is not required during the steam.

generator replacement eutage.

Pse AV; ;R V7ST.T.T--;Y TJPTTT 1 Amendment No.

REACTOR GGGLANT SYSTEMI ClTT'MlT='rT T 7'hTel=

v b.

T, --

r)VnTTrT7)=lkA=?,TlrQ If the recults of the inservic U

ff- vt =

4 --

r==r AI llr1 U Proposed changes to draft page from Unit 1 LAR No. 320 inspection of a stcam 1

generator conducted in aeeerdance with Table 4.4 2 fall into Category C 3, the inspection frequency shall bc increased to at least once per 20 months. The increasc in inspection frequency shall apply until the subsequent inspections satisfy the criteria of specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months.

c. Additional, unscheduled inscrvicc inspections shall be perfer-med en eaeh steam generatre in aeeer-danee with the firot sample inApctien Apecified in Tabl 44A 2 du-ring the s hu t d i-m osubsequent to any ef the f.llwing conditions:

I. rimary to scondary tube loales (net including leaks, originating fr4m tube o , tube so ht welds) in emeess of the limito of Speeifieatien 2.4.6.2,

2. A seismie eeeu~rrenee greater than the Operating Basis Bar-hquaker,

. A o olfs eoolant aeeident rquiring artuatien cf the engineered safeoguards, o-r 4

A 7. -, - - -. -a- -%m

- - a -ar - a- 4- __ 1 4 - - I-.r __ Ir-

21. ~ 2 It U2eIJeaI12 t ier

a. As uscd in this Gpecification:

1. Imperfecetion meano an mceeption to the dimonoionc, finich or contour of a tube from that required by fabrication drawings or peocifications. Eddy currcnt tooting indications blw b 20 pereent of the nominal tube wall thiekneoo, if detoctablo, may be eenoiderod as imperfeetiens.

2. Depradation means a scrvicc induced craceking, wastage, wear or general corroeion occurring on either insidc or outoidc of a tube.

3. Dograded Tube means a tube containing imper-fetiens groator than or equal to 20 percent of the nominal wall thickncs causcd by degradation.

L P er-cnt De-radation means the percentage of the tube wall thicekncs affceted or removed by degradation.

Tn"'7T"?T I¶77 7 Tt " ? T"T¶I"' S

REAC-T-OR OGOLANT SYSTEM Proposed changes to draft page from Unit 1 LAR No. 320 CURVE i LmACEB IZE UI_ 'vrgj-E -1nF nhc.. I1

'. Dfecet mcans an imperfcetion of zuch severity that it ocmceds the plugging limit. A tube containing a defcet is defcetive. Any tubc which dooe not permit thc passage cf the eddy currcnt inspection probc shall be deemed a defective tube.

C. PluqeEincq Limit means the imperfecetion depth at or beyond which the tube shall be removed from zervice by plugging because it may become unservicable prior to the ncxt inspection. Thc plugging limit io equal te the 4O percent of the neminal tube wall thiekneso.

7. Unscrviccablc describes thc condition of a tube if it leaks er eentamin an defet large eneugh to affct_ ito otructural integrity in the event of an Operating Basis Earthquake, a loss of coolant accident, or a oteamline or foedwater line broeak as specified in 4.. S. B. e, above.

8. Tube Inspection means an inspection of the steam genorator tube from the point of entry (hot log oide) completely around thc U bend to the top support of the cold leg.

TTK'ITrV BEAVER VALLEY I hIIAodntI. 34 4 !Gb Amendment Ne.

fl~dL~2.L1ft L~UuL:Uei'r b §.I.LI" Proposed changes to draft page from Unit 1 LAR No. 320 SURYEILLANGE REQUIREMENTS (Cont inued-)

b. The eteam gner-ater shall be Elter-mined GPERU'TLE aft-er empleting the _orros'pnding aetiens (plug all tubes emeeeding the plugging limit) required by Table 4.4 2.

4.4.5.5 foot

a. Within 15 days fellewing the oenpletien ef eaeh --n-------

inspeetien of steam generater tubes, the niurerib ef tubea plugged in eaeh steam gener-ater- shall be submitted in a Speeial Repert in accor-dance with 1:0 CER 50.4.

b. The eemplete results ef the steamn generator tube inser.vi-inspectien shall be submitted in a Speeial Repert in aecor-dance with 1:0 GFR 50.4 within -12 months follewing the completien of the inspectien. This Speeial Repeort shall inelue.e 1:. NIumber and extent of tubes inspected.

2. Loeation and peorcent of wall thiekness penetration fer eaeh indieation of an imperfectien.

3. identifieatien ef tubes plugged.

e. Results of oteamn gener-ater tube inspeetiens whieh faill into Categery C 3 shall be reperted to the CoW......-ocion pursuant to Speeifieatien 6.6 prior to resumptien ef plant-eperatien. The written repert shall provide a deseriptien of investigations eondueted to determine the eause of the tube degradation and eerrective measurec taken to prevent reeurrenee.

BEAVER VALLEY T UNIT 1 TAITT 3 4 lGe

Proposed changes to TABLE 1.1 1 draft page from Unit 1 LAR No. 320 MINIMUM MUMBER OF STEAM CENERATORS TO BE TNrMEC!TDPr DTTRIMTNP TIM£ERVICT XNqP=OTTC)

Preosrvic Inspeetion Ne Yee No. of Steam Ccncrators por Unit Three Three First Inorvicc Inspeetion A4_ Two

£ccond ' Subsequent Inocrvicc ins ectiens OnG1 nc (2)

Table Notation:

(1) The inocrvicc inspection may be limited to onc steam generator on a rotating ochedulo cncompa-oing 9 perocnt of the tubeo if the reoulto of thc firot or previouo inopeetione indicatc that all steam gencratoro arc performing in a licc manncr. Notc that under oome eiroumstanees, the operating conditions in onc or mor soteam generators may be found to bc more scvere than thoes in other steam generatoro. Under such circumstanecs th-csample osquenec ohall be modified to inspeet the moot severe ecnditions.

(2) Thc other otcam qcncrator not inopeeted during_--s the first in-rcrvioo in-ro- on

, vas_ hall_ YAP _

Ie muL _f n____ s s _ s s

- ,, 1-11 4

s

- - - - . I'--- - - - - - - - - -- - , - , , - . 4 - - -, , - +- -i -- - 4 - -

t:J - - ;

,- -_- - ts -\\-ks i X l v-s -- v X s v l l vff vss l si_ X l W l vs;J

/. % _ s

,- / - .e Th T~nt vr7T T T hrrm-*Im sI

_1 _l 3J414 -1d Amendment Ne.

TABLE 4. 21 Proposed changes to draft page from Unit 1 ELMSAM USNERAU^TUR TUBS LiNSLEcTIOfl LAR No. 320 r

I lvl -

.'Y' ^'

vi-L Sffn9

¶A y

-I.

7 fn b-i m T fT h lU Action Required I . -r.-.7 1a n - -A - -T - - TA-.-T

'PiD bfH1L'LS lPibl'^E'l-lUPi Action Required 3Pl AAMPTPr TNAFrECTTOHN Aetien Required Nee Pene defeetive tubes anal Plug defeetive inopect additional 22 tubes and inpAct add itienal tubco in thio G.G. 42 tubog in this A.A. t.bes Perform action fer G 3 result of first Samele P.vrform action for C :

result of first samplf~

Trvno r'l all t41 in Nene thio S.C., plug defcetive tubcs and inspect 2S tubeo in cach other B.C.

Perform action for C :

NA0tif-iRationq toA N7r- r-oult Af seeend s*amo purouant te Specification 6.6 Inopcet all tubes in cach 2.G. and plug defeetive tubes. Notifieation to NRG nursuant to Sreeifieatien

_ n O_ .. I I

.the o "# 9 encre n fl1 numner of oteam generatoro inspeeted dIuring an inspection.

n A rT'1.n ¶77VYr TrrI7 ?Th~TT I Amendment Ne.

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 SG tube intearity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

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

ACTION:

-- - - - - - - - - - - GENERAL NOTE - - - - - - - - - - - - -

Separate action statement entry is allowed for each SG tube.

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

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

2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering MODE 4 following the next refueling outage or SG tube inspection.

b. With Action a not being completed within the specified completion time, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with, and at a frequency specified in the Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Proaram Prior to enterina MODE 4 following a SG tube inspection.

BEAVER VALLEY - UNIT 1 3 /4 4-8 Amendment No.

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 150 gallons per day primary-_=to- secondary LEAKAGE through any one steam generator, and

d. 10 gpm identified LEAKAGE.

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

ACTION:

ba. With any Reactor Coolant System operational LEAKAGE greater than any one of the above not within limits, eenluding for reasons other than pressure boundary LEAKAGE or Primary to secondary LEAKAGE, reduce the LEAKAGE rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STAUDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SCIUTDOWN within the following 30 heou*s.

ab. With the required action and associated completion time of Action a not met. or with e*Py-pressure boundary LEAKAGE.Lor with primary to secondary leakage not within limit, 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 COLD SHUTDOWN within the nuemt following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System operational LEAKAGES shall be I demonstrated to be within each of the above limits by:

a. Monitoring the following leakage detection instrumentation at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:(1)

1. Containment atmosphere gaseous radioactivity monitor.

(1) Only on leakage detection instrumentation required by LCO 3.4.6.1.

BEAVER VALLEY - UNIT 1 3/4 4 -13 Amendment No. 1--C4 I

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued)

2. Containment atmosphere particulate radioactivity monitor.

3. Containment sump discharge flow monitor.

4. Containment sump narrow range level monitor.

b. Performance of a Reactor Coolant System water inventory balance at(2)j(ast once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state eperatien.

c. Verifying primary to secondary LEAKAGE is less than or equal to 150 gallons per day thi 9 ugh any one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

(2) Not required to be performed in rMODE a or 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

(3) Not applicable to primary to secondary LEAKAGE.

BEAVER VALLEY - UNIT 1 3/4 4-14 Amendment No. gr&g- I

ADMINISTRATIVE CONTROLS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (Continued)

The methodology listed in WCAP-14040-NP-A was used with two exceptions:

a) Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1", and b) Use of methodology of the 1996 version of ASME Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure".

c. The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

6.9.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection Derformed in accordance with the Soecification 6.19. Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG.

b. Active degradation (as defined in EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines") mechanisms found.

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications.

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

BEAVER VALLEY - UNIT 1 6-21 Amendment No. 2-Gc I (next page is 6-23)

ADMINISTRATIVE CONTROLS Containment Leakage Rate Testing Program (Continued)

b. Air Lock testing acceptance criteria and required action are as stated in Specification 3.6.1.3 titled "Containment Air Locks."

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

6.18 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 6.18.b.1 & 2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

6.19 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions For Condition Monitoring Assessments Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the BEAVER VALLEY - UNIT 1 6-26 Amendment No. 2-34 l

ADMINISTRATIVE CONTROLS Steam Generator Program (Continued) condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means. prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.

b. Provisions for Performance Criteria For SG Tube Integrity SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity.

accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup. operation in the power range, hot standby. and cool 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 'safetv factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

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

Leakage is not to exceed the operational leakage limit listed in LCO 3.4.6.2.c of 150 gpd per SG.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2."

c. Provisions For SG Tube Repair Criteria Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

BEAVER VALLEY - UNIT 1 6-27 Amendment No.

ADMINISTRATIVE CONTROLS Steam Generator Program (Continued)

d. Provisions For SG Tube Inspections Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not Part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope. inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspcznt-ion met-hods -need to him imm-nlnum aTnd at- what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144. 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE BEAVER VALLEY - UNIT 1 6-28 Amendment No.

Attachment A-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Changes License Amendment Request No. 196 The following is a list of the affected pages:

V XIV xv 1-3 3/4 4-11 3/4 4-12 3/4 4-13 3/4 4-14 3/4 4-14a 3/4 4-14b 3/4 4-14c 3/4 4-14d 3/4 4-14e 3/4 4-14f 3/4 4-15 3/4 4-16 3/4 4-19 3/4 4-20 6-22 6-22a*

6-27 6-28*

6-29*

6-30*

6-31*

6-32*

New page

INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3.3.5 Remote Shutdown Instrumentation .............. 3/4 3-52 3/4.3.3.8 Accident Monitoring Instrumentation .......... 3/4 3-57 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4.4.1.1 Normal Operation ............................. 3/4 4-1 3/4.4.1.2 Hot Standby .................................. 3/4 4-2 3/4.4.1.3 Shutdown ..................................... 3/4 4-3 3/4.4.1.4 1 Loop Isolation Valves - Operating ............ 3/4 4-5 3/4.4 .1.5 Isolated Loop Startup ........................ 3/4 4-6 3/4.4.3 SAFETY VALVES ................................ 3/4 4-9 3/4.4.4 PRESSURIZER.................................. 3/4 4-10 3/4.4.5 STEAM GENERATORS (SG) Tube Integritv ......... 3/4 4-11 I 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 Leakage Detection Instrumentation ............ 3/4 4-17 3/4.4.6.2 Operational Leakage .......................... 3/4 4-19 3/4.4.6.3 Pressure Isolation Valves.................... 3/4 4-21 3/4.4.8 SPECIFIC ACTIVITY............................ 3/4 4-27 3/4.4.9 PRESSURE/TEMPERATURE LIMITS 3/4.4.9.1 Reactor Coolant System ....................... 3/4 4-30 BEAVER VALLEY - UNIT 2 V Amendment No. 43 I Corrected by letter dated July 11, 2002. I

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.3 FACILITY STAFF QUALIFICATIONS ..................... 6-6 6.4 DELETED 6.5 DELETED 6.6 REPORTABLE EVENT ACTION ........................... 6-6 6.7 DELETED 6.8 PROCEDURES ........................................ 6-7 6.9 REPORTING REQUIREMENTS 6.9.1 DELETED 6.9.2 Annual Radiological Environmental Operating Report ......................... 6-18 6.9.3 Annual Radioactive Effluent Release Report ................................... 6-18 6.9.4 DELETED 6.9.5 Core Operating Limits Report .6-19 6.9.6 Pressure and Temperature Limits Report (PTLR).6-21 6.9.7 Steam Generator Tube Inspection Report. 6-22 6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM ..................... 6-22a BEAVER VALLEY - UNIT 2 XIV Amendment No. 118 l

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.12 HIGH RADIATION AREA .............................. 6-22a 6.13 PROCESS CONTROL PROGRAM (PCP) .................... 6-24 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)........... 6-25 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid. Gaseous and Solid) .... .......... 6-25 6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM ......... 6-25 6.18 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM .......................................... 6-26 6.19 Steam Generator (SG) Program ..................... 6-27 BEAVER VALLEY - UNIT 2 XV Amendment No. 44& 1

DEFINITIONS CORE ALTERATION 1.12 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

LEAKAGE 1.14 LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be Pressure Boundary LEAKAGE, or

3. Reactor Coolant System LEAKAGE through a steam generator to the secondary system (primary to secondary LEAKAGE).

b. Unidentified LEAKAGE Unidentified LEAKAGE shall be all LEAKAGE (except reactor coolant pump seal water injection or leakoff) that is not Identified LEAKAGE.

c. Pressure Boundary LEAKAGE Pressure Boundary LEAKAGE shall be LEAKAGE (except Steam generator tube primary to secondary LEAKAGE) through a nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

BEAVER VALLEY - UNIT 2 1-3 Amendment No. 97

REACTOR COOLAN~T SYSTEM 3/4) .. STEAM

£ G7BNBRAT-GR£ LIMITINC CONDITION FOR OPERATIOG 2.4.5 Each steam generator ohall bc OPERABLE.

APPLICABILITYT. IEIDES 1, 2, 3 and 4.

With ene er mere steam g1neraters ineperable, restere the inep7rable gencrator(o) to OPERABLE status prior to increasing T abovo 2000 F.

SURVEILLACE REQUIREMENTS 4.1. .1 £tcam CGncrator .amp.c ciection and Inspcetion Each steam gencrator shall be determined OPERBLE during shutdown by sclecting and inspecting at least the minimum number of steam generators opecified in Tablo 4.4 1.

4.1.5.2 Steam Cenerator Tu:be Samnlcl Seleetion and insigeetien The ateam generater tube mianimffum sample size, inspeetien result-elassifieatien, and the eor-r-pending aetion reqired shall be as apecified in Table 4.4 2. The inscrvicc inspection of steam generator tubes shall be performed at the frequencics specified in Specification 4.4.5.3 and the inspected tubes shall be verified aceeptable per the aceeptance criteria of Specification 4.4.5.4. Steam generator tubes shall be emamined in accordance with Articlc 8 of ecetion V ("Eddy Currcnt Examination of Tubular Producte" ) and Appendim IV to ecetion XI ("Eddy Currcnt Emamination of Nonferromagnetic Steam CGncrator IHeat Exchanger Tubing") of the applicable ycar and addenda of the ASME Beiler and Prcssure Vcsscl Codc required by 10CFR5O, ecetion 90.55a(g). When applying the eceeptions of 4.4.5.2.a through 4.4.5.2.e, previeus defects or imperfectiens in the area repaiodby sleeving are not considered an ao ourn onpcin The tubesz celected for eaeh inservie_ Inspectien shall inlude at least 3 percent: of the toetal number of tubes in all steam generaters; the tubec selected for these inspeetienc shall be colected on a random basis emet

a. Where _ per-ene in similar plants with similar water chemisatry indicatc critical areas to be inspected, then at least 50 pereent of the tubes inspected shall be from thesco critical areas.

b. The fir-et sample of tubes selected for eaeh ic~c inspeetien (subsequent to the preserviee inspeetion) of eaeh steam generater shall inelude-

_ -L4A l _

REACTOR COOLANT CYGTEr?

SuRVEILLANGE REQUIREPIENTS (CGntinued)

1. Al! nenplugged tubcs that previeusly had detectable wall penetration greater than 20 perecnt, and

2. Tubes in thoec areas where eaiperienec has indicated petential preblems, anRd

3. At lcast 3 perecnt of the total number of sleaecd tubes in all thrAe staam g4nerater_. A sample isa-lc03 than 3 perecnt is aeeeptable previded all the sleeved tubes in the steaam generater(s) ea2fained during the refueling eutage arc inspected. Thcse in4peetiens will inluda b4th the tube and the sleeve-A a-nd

1. A tub inspeetiefl pursuant te Speeifieatien 1.1...a.8. If any sclected tube doc net permit thc passage ef the eddy eurrent prbe f.or a tube or sclacv inspectien, this shall bc recrded and an adjacent tuba shall be sclected and subjected to a tuba incalatien.

5 S. indieatiens left in erviee as a result Cf applieation.

of the tube supp art plate veltage based repair

__4 r4 _-

wrlecrle A

If1 A

^+

r7 b +A

. d_

10) shall ba inopected by bobbin eeil proeba during all future refueling Cutagac.

_ L _ fI A 1I

c. nhe tubce scleacad as thC sccna ana tnira dsamples toar reuired by Table 44A 2) during aah insarviea in peti4n may be subjected to a partial tuba inspectien previded.

1. The tubea salected for thecs samplcs includa the tubc from thecs areas Cf the tuba sheet array where tubcs with imperfeetiens were previeu ly feund, and-

2. The inpatin inluda A thse C .

_bas ' cs whera imperfeceti Lons- wer previeusly feund.

d. Implementation Cf the stcam generator tuba to tuba support platc repair critcria requircs a 100 perecnt bobbin coil inspectien for hot leg and cold leg tuba suppert plate interscetiens down to the lewect cold leg tuba support plata with ]cnown Cutsida diameter strcss corrosion cracking (OD£CC) indication3. The determinatien Cf the lowect cold leg tuba aupport plate intersectiens having OD^CC indications shall ba based en the perfermance Cf at least a 20 perecnt random sampling Cf tubes inspected ever their full length-.

bbl kVL1t VfiL3LLY fi+/-: Ue.

l Amendment lie. 101

tftbil0=+/-ldtt §=Uutdttft bltb+/-:bt,.m GURYEILLANGCE R-EQUIREPIENTS (Gentinued)

Thc rcsults of each sample inspection shall b colassified into onc of -

thc following threc eategeries;

+/-Rspeeea:aR ftestt+/-ea Lcss than ' pereent of thc tetal tubec inspected are degrade'

-- , L , - - -rA -3 -- - -C -t - --

-. .4.4.tj dtttt' ttt'.fli' '.4. ttt 4.*Jvivfr'..

tubes are defeetive.

One er mere tubes, but net mere than 1 pereent ef the total tuben ins~pected are defeetive, or l ,,^\ ._ V- , . -. _wP

-1 ..- -

ef the tetal tubc inspected are degraded tubes.

MPrc than 10 perecnt of thc total tubc inspected are degraded tubeos r freo than 1: perent Cf the inspected tubes are defeetiveo.

Motc; In all inspectiens, previeusly degraded tubes er sleeves mnost C2mhibit si~gnif-iean~t (greater than 10 perrcnt) further wall penCtratiens te be ineluded in the abeve 4.41.S.3 Insoection Fr-euencios The above required inscrvicc in:petiens Cf steam generato-^ tubsc shall be perfermed at the fellewing founio

a. The first inscrvicc in pectien shall bc perfermcd aftcr

- T S==- ___l 7 s_ - T ._ ~1_

_-t c LfeOtlvc Full eower reonenc but witnin 2 calenaar montnh Cf initial criticality. Subscquent inscrvico inspections shall bc performod at intcrvals Cf net less than 12 nor morc than 24 calendar months after the previeus inspectien.

If two cnsecutive inspectiens following ccrvicc under All Velatile Trcatmcnt (AVT) conditions, net including thc prcscrvicc inspection, result in all inspectien rcsults falling into thc C 1 category Cr if two cnsecutivc inspectiens demonstrate that previeusly observed degradation has not continued and no additienal degradatien has ccurred, thc inspectien intcrval may bc oetended to a mArmirm of M nf- r'-r 10 A month^'

T "AI T"lT r '7kT T "'XP TThTTM 'I J- / 4- 4 -- Amcnamcnt Neo. +/-1U

REACTOR COOUANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

b. If the inscrvic inspection of a stcam gencrator conducted in accordance with Table 4.4 2 requirce a third samplc inspection whozs results fall in Category C 3, the inspection frequency shall be inorcased to at lcast once per 20 months. Thc incrcasc in inspection frequency shall apply until a subsequent inspection demonstratcs that a third sample inspection is not required.

c. Additional, unscheduled inscrvicc inspections shall be performed on cach steam gencrator in accordance with thc first Sample inspection specified in Table 4.4 2 during the chutdown cubscquent to any of the following conditions-

1. Primary to sccondary tube leak] (not including leaka originating from tube to tubecsheet welds) in comecs of the limits of Specification 3A.4..2,

2. A ecismic occurrcnce greater than thc Operating BasiB Earthquae, B. A less of eoolant aeeident requiring aetuation ef the nd cafeguards, or

-. A main etcamlinc or foedwater line break.

LL45S. Aeeeptanee Criteria'

a. As uscd in this Gpecification;

1. Imnerfcetien means an coeeption to the dimcnsionz, finich cr contour of a tube or sleeve from that required by fabrication drawings or specificationo.

Eddy currcnt testing indications below 20 perecnt of thc nominal tube wall thicknces, if detectable, may bh considered as imperfectiono.

2. Dqradatien means a _ervice indueed cracking, wastage, wear or general corrocio- occurring on either inside er ouAtsiide of a tube or sleeve.

B. Doraded Tube moans a tube or sleve containing imperfeetiens greater than or equal to 20 percent of the nominal wall thiekness eaused by degradation.

M.; '; -Iz _

3L 4 4mndment ie. :EVui

CURVEILLANT E REQUIREMENTS (Gentinu4d)

4. Perecnt Dcqradation mfans the percentage of the tube or sleeve wall thicekncss affceted or removed by degradatioen.

5. Dfecet moans an imperfcetion of such severity that it coeeeds the plugging or repair limit. A tubc containing a defcet is defcetive. Any tube which dooc not permit the passage of the eddy currcnt inspection probc shall be deemed a defcetive tube.

6. Plugging or Rcpair Limit means the imperfecetion depth at er beyend whieh the tube shall be reomoved frem scrvico by plugging or repaired by sleeving in the affected area bcaue -it may becomo unse.v_-eabl-prior to the noect inspection. The plugging or repair limit imperfecetion depths arc opecified in percentage of neminal wall thickness as follews.

a) Original tube wall 40° This definition docs not apply to tube support plate interseetiens for which the voltzage based repair criteria arc being applied. Refeor to 1.4.5.1.a.10 for the repair limit applicable to these nosotos b) AD Combustin Engineering TIC welded 32_V-sleeve wall c) Westinghouse laser welded sleeve wall 25°

7. Unserviceable describes the condition of a tube if it leaks or cont-ain;:s am defect large enough to affect its structural integrity in the event of an Operating nasis _artheake, a less of coolant aeeident, or a steamline or foedwater line break as specified in 1.4.6.3.c, above.

3. Tubc Inspection moans an inspection of the steam geneirater tube from the point of ento- (hot leg side-)

completely around the U bend to the top support to the eed4egr T) T' M T '"n XT TT T T'Av TIThTm IN 3I4 4 l4a Amendment No. 101

REACTOR GOOL NT SYSTEM CURVEILLANCE REQUIREMENTS (C.ntinuaed)

9. Tube Repair refers to sleeving which is used to maintain a tube in serviee or return a tube to servioe. Thi3 ineludes the removal of plugs that were ins ta __Ad a.s a .erreetive er prevAntive maAur. ThA following sleeve designs have been found aceeptabla.

a) ABB Gmbustion EngineAA-Ing T-IC wAlded sleeve-A, CEN 629 P, Revisien 02 and CEN 629 P Addendum 1.

I, I~ _,I- rs . - fl o b) Westinghouse laser wejlaea elee-vZe, 1-1-~

vik=1t Revisien 1.

10. Tube Support Plate Plugging Limit is used for the dispesitien of an alloy 6OO steam ganarator tube for eontinued scrvicc that is eaxperianeing predominantly axially eriAntad AutAsid diamAter stress _orroci^n eracking confined within the thicknecs of the tuba suppert plates. At tuba suppert plate interseetiens, t1aidplugging ~reapair-) limit: is baseet on maintaining steam generator tuba scrviceability as described belew:~

a) Steam genarator tubcs, whoec degradation ic attributed t outsAidA diamAter stress AArresion_

cracking within the bounds of the tuba support plate with b1bbin veltages lecs than or equal to 2.0 volts will be allowed to remain in scrvicc.

b) CSteam _eneratr tubAs, whose dAgradation i attributed to outside diameter strcss corrocion craceing within the bounds of the tuba support platC with a bobbin veltage greater than 2.0 volts will be repaired or plugged, mcecpt as netad in A.A .5..a._Io.e balow.

i-i ' V'I-I VI#LL U 1/-- 34 4 4 4:4b Amnmat 149.

Affiendment No 0 1

REACT-GR? COOGL2AT- CYSwETEM SUTVEILLA~NGE REQUIRBMENTS (Centinued) e) Steam generater tubes, with indieatiens cf petential degradatien attributed te eutside diameter stress eorroezien eraeking within the beunds ef the tube cuapiprt; Plate with a bebbin vltage greater than 2. 0 velts but less than er equ l to the upper veltage roar iit-ay remain in servicoe if a rotating paneake eoil or acceptable alternative inspeetien dees no..t deteeotr degradat-ion.- Steam generater tubes, with indieations of eutside diameter stress -- rr----n eraeking dgradatin with a bobbin t _ greateor than the upper voltago repair limit will bo pluggtd or rtepairld.

d) If _n unocheduled mid eyele inspeetion-fl ic perfermed, thc following mid cycl L repair limits apply A I - A ini3tead

- . - - A of A-the A

limits

n- 10 _ Al .

identified I A. A__

in n

The mid cyclc repair limits ,arcdw cmip from thcp following cqu a - --- n-:

VSL IV1u 1.O+NDE+Gr (ICLi )

CL

'7 7j~I r )(CL-At)

VMLRL -'MURL ('URL-VLRL( CLC-L (1) The upper veltago repair- limit is ealeulated aecording to the fnethedolegy in Cenoric Letter 95 05 as supplemented.

BEAVER VALLEY tRIT 2 3,'1 4 llc Amendment No. 101

RSEACTOR COOLGANT SYSTEM GURVB-TET3TLE3GnREQUIREMENTS (Gentinued) upp.r vltage repaair 4 limit lower voltage repair 1iifni t limid tyle upper veltag repair limit based en time into eyele

'mid cycl loewer veltage repair limnie basd oR Vt+ and timc into cyel At length of time since last scheduled inspection during which

.4-5HJJ RLwe-re implementod CL eielc length (the time between two scheduled stcam generator inspections)

-truetural limit voltage Cr- average growth rate per cyele length NDE 95 perecnt cumulative probability allowanee for nondostructive examination unecrtainty (i.e., a valuc of 20 pereont has been Implementatien of thcse mid cyclc repair limits should fellew the same appreaeh as in TS A.4.SAa.!.a, 4.4S.4.a. .b, and .4.5..a.l0.e.

(2) The !ME is the valuc prcvided by the NRc in CL 95 05 as supplemented.

_.s . v os 1 vs _ w 3,1 4 ^14 LZIPmn mnY

REACTOR GGGL7NT SYCTEM GURVEILLAICE REQUIREMENTS (Continued)

b. The stcam generator shall be determined OPERABLE aftar eempleting the corrcsponding actions (plug or repair all tubes emeeeding the plugging or repair- lifmit) r-equired by Tablo 4.- 2.

L4.5.5~ flpe<orto

a. Within 15 days follewing the eompletien of eaehinric inspetien ef soteam genr-atr tubes, the numnber of tube_

plugged er repaired in each steam generator shall be oubmittad in a Speeial Repert in accordancA with 10 GRP

b. The complete rcoults of the steam generator tube and sleeve inscrvic inspection shall be submitted in a Special Report in accordance with 10 CFR 50.4 within 12 months following thc completion of the inspection. Thic Special Report ohall includc;

1. Number and extent of tubes and sleeves inspected.

2. Location and percent of wall thiccncss penetration for cach indication of an imperfecetion.

B. identifieation of tubes plugged or repaired.

c. Results of steam gnerator tube inspeetiens whieh fall intv Catcgery C 3 shall be reperted to the Commiccion pursuant to Gpeeificatien 6.6 Prior to reAumptiAn of plant eperatien. The written repert shall previda a deseription of inveotigationo eendueted to determine the eause of the tube degradation and eorreetive measures taken tc prevent reeurronee.

d. FPr impl.mhntation of thA voltage based ropair _ritria to tuba suppert plate interseetiens, notify the Commission prior to returning the steam generatora to eorvice (?4ODE 4) should any of the following conditions ariocE~.-

1. If estimated lealsage based on the prejected end of cyc 10 (or if not practical, using the actual measured and of cyelA) voltago distribution _ ds the leak limit (determined from the licvnsing basis dosv caleulation for the pectulatmd main staamlint breok) for the nrxt opcrating eycat.

AVL10 VALL H Y UN 1'1 3/4 4 llc Amendment No. 101

REAGCT. COOGLANT CSCTEM SURVELLVANE REQUI REMETS (Cent inue d)

2. If circumferential crack lise indications arc detceted at thc tube suppert plate intersections.

3. If indications arc identified that emtend beyend the confincs of the tube zupport platc.

4. If indicatiens arc identified at the tube zupport platc elevations that arc attributable to primary watcr strcss corrosion cracking.

'. If the calculated cnditienal burst probability bascd on the prejected end of cyclc (er if net practieal, using the actual measured end of cyclc) voltagc distributien emv ds 1 X -1O n tify thea Gmmissien and previda an asaessment ef the safety signifieanee ef the eeeurrenee.

flT'7DrTT Ln !T7kT T T - TThTTrT 3B4 14lf k- #

A ; pndmcnt 7 . nf%

101

MIUITtTM NTJBIBLR OF STEAM GEUENRATORS TO BE iNGPBCTBD DUTiNC RiHERViGCE lIIPECTION Prccrvicc Inspection -le Yes No. of Stcam Ccncrators per Unit Three -Three First Inrvio:-c inpscc^tin A1 - T-w Sccond & Subscuent Inocrvicc Inspections Gne -GneG Tablc Notation I. T-he inserviee inspecctien may be limited te ene steam generator~

on a retating oehedule eno one 9^ of^ the tubes if the rcults of thc first or previous inspeetions indicate that all otcam gencrators arc performing in a likc manner. Notc that under somc eircumstances, thc operating conditions in onc or morc steam gencrators may bc found to bc morc sevcrc than thosc in other steam gencrators. Under such cireumstanecs thc samplc sequencc shall bc modified to inspect thc most oevcrc conditions.

2. Thc ether stcam gencrator not inspected during thc first inscrvice inspection shall bc inspected. Thc third and oubseeuent inspctions should follow thc instruction described T'r.TAVt.'r. . I

-LEVt 7TT 'rT'V Tr TlKTIT, "

-15 urs "I

- .44b- a /aM

TABLE; 4.1

&TEM CEGElRATOR TUBB INSPECTION lST SAMPLE ITGPBCTIOUG RID SAMPLE I?;PBGCTIUGN 3RD SAMPLE INPEGCTIOII 6amplci-a-6e Result Action REquired esult Action fRquired Reseid Action Required A minimum G__I_ Nene N74A /A N/IA/A Of C tubez Q Plug or repair Nene N/A N/A per G.G. defeetie_

tubco and inopeet C 2 Plug or repair defcetive ____ Netne additional 2S tubes in. tubcs and inspeet-additional C 2 Plug or repair thio S.C.- 1S tubea in this £.C. defoetive tubce C-3 Prrform action for C 3 rcoult of4 firot oamplc C 3 Pcrform action for C 3 N/A .-

reoult of firot oamplc_

C Inopeet all tubco in All ether Iene Nl/A N/A thio S.C., plug or G.G.s are repair defoetive tubco G-4 and inopect 22 tubco _

in cach other S.C.- Somc £.C.o Pcrform action for C 2 NA/A C 2 but no reoult of ocoond sample Notification to NRG additional Specification 6.6. G- _ _

Additional Inopect all tubcs in cach N/A N/A G.C. is L.C. and plug or repair C 3- defcetive tubco.

Notification toeRC pur-uant

. to Specification C.G._

- . .. I- - - - __ . - . I I - - - I , - . - - - - __ __ - - - . . - -

U - Whnre n iS the numoor oe seeam gencrators inispetcd during an inorcetion.

-_T" ~VfULLLY

- IdVElt rTT ITETTV UPi1 l '- Amendment No. 101

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 SG tube intearitv, shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.

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

ACTION:

- - - - - - - - - - - - - - - GENERAL NOTE - - - - - - - - - - - - -

Separate action statement entry is allowed for each SG tube.

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

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

2. Plug or repair the affected tube(s) in accordance with the Steam Generator Program prior to entering MODE 4 following the next refueling outage or SG tube

.,inspection.

b. With Action a not being completed within the specified completion time, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.5.1 Verify SG tube integrity in accordance with. and at a frequency specified in the Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged or repaired in accordance with the Steam Generator Program prior to entering MODE 4 following a SG tube inspection.

BEAVER VALLEY - UNIT 2 3/A4 4-11 Amendment No.

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 150 gallons per day primary-=to- secondary LEAKAGE through any one steam generator, and

d. 10 gpm identified LEAKAGE.

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

ACTION:

ha. With any Reactor Coolant System operational LEAKAGE greater than any one of the above not within limits, exeludi-ng for reasons other then pressure boundary LEAKAGE or primary to secondary LEAKAGE, reduce the LEAKAGE rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least !IOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD CIIUTDWGU; within the following 30 heurs.

ab. With the required action and associated completion time of Action a not met, or with any-pressure boundary LEAKAGE, or with primary to secondary leakage not within limit, 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 COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System operational LEAKAGES shall be demonstrated to be within each of the above limits by:

a. Monitoring the following leakage detection instrumentation at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:(1)

1. Containment atmosphere gaseous radioactivity monitor.

(1) Only on leakage detection instrumentation required by LCO 3.4.6.1.

BEAVER VALLEY - UNIT 2 3/4 4-19 Amendment No. God1 I

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS (Continued)

2. Containment atmosphere particulate radioactivity monitor.

3. Containment sump discharge flow monitor.

4. Containment sump narrow range level monitor.

b. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state eperatien. (2)(3)

c. Verifying primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .(2)

(2) Not required to be performed in MIODE 3 or 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

(3) Not applicable to primary to secondary LEAKAGE.

BEAVER VALLEY - UNIT 2 3/4 4-20 Amendment No. G4& l

ADMINISTRATIVE CONTROLS PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

c. The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

6.9.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.19, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG including number and extent of sleeves examined.

b. Active degradation (as defined in EPRI. "Pressurized Water Reactor Steam Generator Examination Guidelines") mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged or repaired to date.

g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

h. The effective plugging percentage for all plugging and tube repairs in each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

For implementation of the voltage-based repair criteria to tube support plate intersections, notify the Commission prior to returning the steam generators to service (MODE 4) should any of the following conditions arise:

1. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.

2. If circumferential crack-like indications are detected at the tube support plate intersections.

BEAVER VALLEY - UNIT 2 6-22 Amendment No. 44a

ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (continued)

3. If indications are identified that extend beyond the confines of the tube support plate.

4. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2. notify the Commission and provide an assessment of the safety significance of the occurrence.

6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.1601 of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shl be controlled by requiring issuance of a Radiological Work Permit '. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

(1) Radiation protection personnel, or personnel escorted by radiation protection personnel in accordance with approved emergency procedures, shall be exempt from the RWP issuance requirement during the performance of their radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.

BEAVER VALLEY - UNIT 2 6-22a Amendment No. I

ADMINISTRATIVE CONTROLS TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM (Continued)

2. a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 6.18.b.1 & 2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

6.19 STEAM GENERATOR (SG) PROGRAM A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition. the Steam Generator Program shall include the following provisions:

a. Provisions For Condition Monitoring Assessments Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means. prior to the plugging or repair of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.

b. Provisions for Performance Criteria For SG Tube Integrity SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity.

accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup. operation in the power range, hot standby, and cool 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 BEAVER VALLEY - UNIT 2 6-27 Amendment No. 12-0 I

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) 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 leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 150 qpd per SG specified in LCO 3.4.6.2.c. except for specific types of degradation at specific locations as described in Technical Specification 6.19.c.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2.

c. Provisions For SG Tube Repair Criteria Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except when alternate tube repair criteria permitted by technical specifications are satisfied.

Sleeves found by inservice inspection to contain flaws with a depth equal to or exceeding the following percentages of the nominal tube wall thickness shall be plugged:

ABB Combustion Engineering TIG welded sleeves 32%

Westinghouse laser welded sleeves 25%

The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1. Tube Support Plate Voltage-Based Repair Criteria (granted by License Amendment 101)

Tube Support Plate Plugging Limit is used for the disposition of an alloy 600 steam generator tube for BEAVER VALLEY - UNIT 2 6 -28 Amendment No.

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:

a) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.

b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in below.

c) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented) may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect degradation.

d) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the upper voltage repair limit (calculated according to the methodologv in Generic Letter 95-05 as supplemented) will be plugged or repaired.

BEAVER VALLEY - UNIT 2 6-29 Amendment No.

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) d) If an unscheduled mid-cycle inspection is performed.

the following mid-cycle repair limits apply instead of the limits specified in 6.19.c.l.a, 6.19.c.l.b and 6.19.c.l.c.

The mid-cycle repair limits are determined from the following equations:

V V= SL L.O+NDE+Gr( CL )At CL VMLRL VMURL -(VURL VLRL)( CL )

where:

Vi = uer voltage repair limit Vn = lower voltage repair limit V = mid-cycle upper voltage repair limit based on time into cycle V<< = mid-cycle lower voltage repair limit Mbased on Va and time into cycle At = length of time since last scheduled inspection during which V.,and._

were implemented CL = cycle length (the time between two scheduled steam generator inspections)

= structural limit voltage Gr = average growth rate per cycle length NDE = 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC) The NDE is the value provided by the NRC in GL 95-05 as supplemented.

d. Provisions For SG Tube Inspections Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting 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 BEAVER VALLEY - UNIT 2 6-30 Amendment No.

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. Previous defects or imperfections in an area repaired by sleeving are not considered an area requiring re-inspection. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment. to determine which inspection methods need to be emnloved and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.

3. If crack indications are found in any SG tube. then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube. diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

4. Inspect at least 3 percent of the total number of sleeved tubes in all three steam generators. A sample size less than 3 percent is acceptable provided all the sleeved tubes in the steam generator(s) examined during the refueling outage are inspected. These inspections will include both the tube and the sleeve.

If any selected sleeve does not permit the passage of the eddy current probe for a sleeve inspection, this shall be recorded and an adjacent sleeve shall be selected and subjected to a sleeve inspection.

BEAVER VALLEY - UNIT 2 6-31 Amendment No.-

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued)

5. Indications left in service as a result of application of the tube support plate voltage-based repair criteria (6.19.c.1) shall be inspected during all future refueling outages.

Implementation of the steam generator tube-to-tube support Plate repair criteria requires a 100-percent inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length.

e. Provisions for monitoring operational primary to secondary LEAKAGE

f. Provisions For SG Tube Repair Methods Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SQ tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

1. ABB Combustion Engineering TIG welded sleeves, CEN-629-P. Revision 02 and CEN-629-P Addendum 1, granted by License Amendment 98 (License Condition 2.C(1-1)

(License Appendix D, Amendment 98) also applies).

2. Westinghouse laser welded sleeves. WCAP-13483, Revision 1. granted by License Amendment 56.

BEAVER VALLEY - UNIT 2 6-32 Amendment No.

Attachment B-I Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Bases Changes License Amendment Request No. 324 The following is a list of the affected pages:

PAGE Pending LAR B-II B 3/4 4-2 B 3/4 4-2a 320 B 3/4 4-3d*

B 3/4 4-3e*

B 3/4 4-3f* 320 B 3/4 4-3g B 3/4 4-3h B3/44-3i B 3/4 4-3j I

Provided for readability only

ProvidedforInformation Only.

TECHNICAL SPECIFICATION BASES INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS ...................... 3/4 4-1 3/4.4.3 SAFETY VALVES .............................. 3/4 4-lg 3/4.4.4 PRESSURIZER ................................ 3/4 4-2 3/4.4.5 STEAM GENERATORS (SG) Tube Integrity. 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .............

I 3/4 4-3 3/4.4.6.1 Leakage Detection Instrumentation .......... 3/4 4-3 3/4.4.6.2 Operational Leakage ........................ 3/4 4-3d 3/4.4 .6.3 Pressure Isolation Valve Leakage ........... 3/4 4-3j 3/4.4.8 SPECIFIC ACTIVITY ......................... 3/4 4-4 3/4 .4.9 PRESSURE/TEMPERATURE LIMITS ............... 3/4 4-5 3/4.4.11 RELIEF VALVES ............................. 3/4 4-11 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS .............................. 3/4 5-1 3/4.5.2 AND 3/4.5.3 ECCS SUBSYSTEMS.................... 3/4 5-la 3/4.5.4 BORON INJECTION SYSTEM .................... 3/4 5-2 3/4.5.5 SEAL INJECTION FLOW. 3/4 5-3 BEAVER VALLEY - UNIT 1 B-II Change No. 1-414-029 l

DPR-66 REACTOR COOLANT SYSTEM I ProvidedforInformation Only.

BASES 3/4.4.3 SAFETY VALVES (Continued)

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.4 PRESSURIZER The requirement that (150)kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

'4 /A A STEAM GENERATORS Onc GEPABLEL sttam generator in a non iselated rcactor coolant loop prevides sufficicnt heat remeval capability to remoev decay heat after a rcactor ohutdown. The rcquircmcnt for two OPEPABLE steam generatrs _embined with other requiroement of the Limitin-Genditiens

-b- q 1 4 rIIA r fer-- Gperation n

_ 4e 4---------- ensures t s~ --

adequate

_ . -A -~r -

deeay

-1__s f-r r.

heat

__ _remeval Ao s0 _- I-- t: tv__vsj = Ctvs _ r _l;Z t= z _: _-r Ur== U_ _ WI = v--_ .tt genrrater beeeme ineperable : due te single failure nsid-ratiens.

_elow 350 0 F, decay heat is r__moved by the RH1 system.

The Cuirveillanee Requirements fr inspectien 4 f the steam generator-tubes ensure that the struetural integrity ef this portion ef the R1CC will be maintained. The program for inservic inspectien of stcam gencrator tubcs is based en a modification of Regulatory Cuide 1.83, Revizion 1. Inscrvic inspection of steam gencrator tubing is cssential in order to maintain ourveillance of the conditions of thc tubes in thc cvent that there is evidenec of mcchanical damage or progrcssivc degradatien due to design, manufacturing crrors, or inscrvicc conditions that lead to corrosion. Inscrvicc inspection of steam gencrator tubing also provides a means of charactcrizing the nature and causc of any tubc degradation so that corrective mcasurcs can be taken.

The plant is expeeted to be eperated i'n a Manner sueh that the seee~ndary eeelant will be maintained within these parameter limit-&

feund te result in negligible cerresien Cf the steam generater tubes.

If the seeendary eeelant ehemistry is net maintained within theoco parameter limits, lecalized eer-r-sien may likely result in streso eerresien crackeing. The extent of crackeing during plant BEAVER VALLEY - UNIT I B 3/4 4-2 Change No. 1-0-01029 I

REACTOR COOLANT SYSTEM I Providedfor Information Only.

Proposed changes to draft page from Unit 1 LAR No. 320.

BASES 9/4.4.S STEAM GENBRATORS (Continued operation would be limited by the limitation of steam generater tube lcakage between the Primary Coolant System and the secondary Coolant System (primary to secondary LEAKACE - 150 gallons per day per steam generato}r) Maintaining a primary to secondary LEMACE less than this limit helps to ensure adequate margin to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that primary to secendary LEAACE ef 150 gallens per day per steam generater can readily be deateted.

Laeakage in carcazs ef this limit will require plant shutdown and an unscheduled inspection, during which the lcaking tubes will be iccated and pluggad-.

Wastage type defcets arc unlikely with proper chemistry ef secondary coolant, such as provided by All Velatile Treatment (AVT). IHewever, even if a defcet of similar t5ae should develop in asrvice, it will be found during scheduled inservica steam generater tube ecaminations. Plugging will be required of all tubes with imperfecetions ecaeeding the plugging limit. Steam generator tube inspections of operating plants have demonstrated the capability to reliably deaet a wastage type defcet that has penetrated 20 perecnt of the original tube wall thickness.

Whenevar the rcsults of any steam generator tubing inc-r-ica inspection fall into Category C 3, these results will be reported to the Commission pursuant to Gpecification 6.6 prior to resumption cf plant operation. Such eases will be considered by the Commission an a ease by ease basis and may result in a requirement for analysis, laboratory examinatiens, tests, additional addy current inspection, and revision of the Tachnical Specifications, if necessary.

3/4.4.5 Steam Generator (SG) Tube Integrity BACKGROUND Steam generator tubes are small diameter, thin walled tubes that carry Primary coolant through the Primary to secondary heat exchangers. The SG tubes have a number of important safety functions.

Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and. as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition. as Part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the Primary system.

This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by "Reactor Coolant Loop" LCOs 3.4.1.1 (MODES 1 and 2). 3.4.1.2 (MODE 3), and 3.4.1.3 (MODES 4 and 5).

ProvidedforInformation Only.

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatorv requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Depending upon materials and design, steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 6.19, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.19, tube integrity is maintained when the SG performance criteria are met.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 6.19. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by NEI 97-06, "Steam Generator Program Guidelines".

APPLICABLE SAFETY ANALYSES The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary SG tube LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.6.2.c, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes that following reactor trip the contaminated secondary fluid is released to the atmosphere via safety valves. Environmental releases before reactor trip are discharged to the main condenser.

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.) In these analyses, the steam discharge to the atmosphere includes primary to secondary SG tube LEAKAGE equivalent to the operational leakage limit of 150 qpd per SG. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EOUIVALENT I-131 is assumed to be equal to the LCO 3.4.8, "RCS Specific Activity," limits. Pre-accident and concurrent iodine spikes are assumed in accordance with applicable regulatory guidance. 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 events are within the limits of 10 CFR 50.67 as supplemented by Regulatory Guide 1.183.

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

LCO I Providedfor Information Only.

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

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still retain tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.19, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall.

The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

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

design basis loads based on ASME Code,Section III, Subsection NB and Draft Regulatory Guide 1.121. "Basis for Plugging Degraded Steam Generator Tubes". Auaust 1976.

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not result in a tube leakage per SG in excess of the operational leakage limit provided in LCO 3.4.6.2.c of 150 gpd per SG. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.6.2, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1. 2, 3. or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1. 2. 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low. resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the actions may be entered independently for each SG tube. This is acceptable because the required actions provide appropriate compensatory actions for each affected SG tube. Complying with the required actions may allow for continued operation, and subsequently affected SG tubes are governed by subsequent condition entry and application of associated required actions.

a. ACTION a applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 4.4.5.1. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to

ProvidedforInformation Only.

determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Action b applies.

A completion time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, ACTION a allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.

However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This completion time is acceptable since operation until the next inspection is supported by the operational assessment.

b. If the required actions and associated completion times of ACTION a are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The allowed completion times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 4.4.5.1 During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, "Steam Generator Program Guidelines", and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program in conjunction with the degradation assessment determines the scope of the inspection and the methods

Providedfor Information Only.

used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e.. which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program and the degradation assessment also specify the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 4.4.5.1. The Frequency is determined by the operational assessment and other limits in EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines". The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition. Specification 6.19 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 4.4.5.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 6.19 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s).

NEI 97-06 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Freguencv of "prior to entering MODE 4 following a SG inspection" ensures that SR 4.4.5.2 has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondarv pressure differential.

BEAVER VALLEY - UNIT 1 B 3/4 4-2a Change No. 1-4G-7029 l

REACTOR COOLANT SYSTEM ProvidedforReadability Only.

BASES 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

SURVEILLANCE REQUIREMENTS (SR)

SR 4.4.6.1.a SR 4.4.6.1.a requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

SR 4.4.6.1.a requires the performance of a CHANNEL FUNCTIONAL TEST on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. The test verifies the alarm setpoint and relative accuracy of the instrument string. The Frequency of 31 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

SR 4.4.6.1.a also requires the performance of a CHANNEL CALIBRATION on the required containment atmosphere radioactivity monitor. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 18 months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this Frequency is acceptable.

SR 4.4.6.1.b SR 4.4.6.1.b requires the performance of .a CHANNEL CALIBRATION on the required containment sump monitor. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 18 months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this Frequency is acceptable.

3/4.4.6.2 OPERATIONAL LEAKAGE BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

BEAVER VALLEY - UNIT 1 B 3/4 4-3d Amendment No. 183 1

REACTOR COOLANT SYSTEM Providedfor Readability Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

BACKGROUND (Continued)

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30, requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area -is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100 percent leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

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 BEAVER VALLEY - UNIT 1 B 3/4 4-3e Amendment No. 183

REACTOR COOLANT SYSTEM I ProvidedforReadability Only.

Proposedchanges to draft page from Unit 1 LAR No. 320.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

APPLICABLE SAFETY ANALYSES (Continued) affect the probability of such an event. The safety analysis for an event resulting- in steam discharge to the atmosphere assumes a 450 gpd (150 gpd per steam generator) primary-to-secondary LEAKAGE.

Primary-to-secondary LEAKAGE is a factor in the dose assessment of accidents or transients that involve secondary steam release to the atmosphere, such as a main steam line break (MSLB), a locked rotor accident (LRA), a Loss of AC Power (LACP), a Control Rod Ejection Accident (CREA) and to a lesser extent, a Steam Generator Tube Rupture (SGTR). The leakage contaminates the secondary fluid. The limit on the primary-to-secondary leakage ensures that the dose contribution at the site boundary from tube leakage following such accidents are limited to appropriate fractions of the 10 CFR 50.67 limit of 25 Rem TEDE as allowable by Regulatory Guide 1.183. The limit on the primary-to-secondary leakage also ensures that the dose contribution from tube leakage in the control room is limited to the 10 CFR 50.67 limit of 5 Rem TEDE. Among all of the analyses that release primary side activity to the environment via tube leakage, the MSLB is of particular concern because the ruptured main steam line provides a pathway to release the primary to secondary leakage directly to the environment without dilution in the secondary fluid.

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 BEAVER VALLEY - UNIT 1 B 3/4 4-3f Change No. 1-027

REACTOR COOLANT SYSTEM Providedfor Information O BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

LCO (Continued) unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Should pressure boundary LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Primary- to- Secondary LEAKAGE through Any One SG Operating eperienec at PWR plants has shown that sudden increases in lea]: rate arc often precursors to larger tube failures. Maintaining an operating LEAKACE limit of 150 gpd per steam generator will minimize the potential for a larga LETTACE event at power. This operating LAKAGCE A limit is more restrictive than the operating LEAKACE limit in standardized technical specifications. This provides additienal margin to aeeommedate a tube flaw whieh might grew at a greater than expeted rate or un_,peoetAdly mAtOnd utsidae the thiekness of the tuba suppert plate. This redueed LEAKAGCE limit, in eenjunetien with a leak rate moenitering program, prevides additional assurance that this prAeurser LEAKAGE will be dataetad and tha plant shut down in a timely mannar-.-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". 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 frecuency of steam generator tube ruptures.

BEAVER VALLEY - UNIT 1 B 3/4 4-3g Amendment Chance No. p-9-1-029

REACTOR COOLANT SYSTEM BASES I

IProvidedfor Information Only. I 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

d. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.6.3, "RCS Pressure Isolation Valve (PIV)," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

ACTIONS ba. Unidentified LEAKAGE--or identified LEAKAGE, or primary to secondary LEAKACGE in excess -of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

If the unidentified LEAKACE, identified LEAKACE, or primary TIV __=tX9CV=--V _rAh Ad _=zSWI___V_ a E

w'ithin 4 houro, tho reacter must be broughto te lEweor S r ~..L

~ '.--- l .A, .L.'.f 4 J L-W-l'.

. . . tEA'.. - b.J- _ 1.,- A LL 1...- T " 1Ar .

and its potential conscquenecs. The reactor must be breught to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE S within -36 heurs. This a.tion rvducco tha LEAKAcE.

The allowed Completion Times are reasonable, based on operating ctperienec, to reaeh the required plant Anditiens from full p.wer eonditins_ in an orderly manner and without challenging plant systems. In ?IODE 5, the pressure stresses aeting on the GCPB arc mueh lower, and further deterioration is mueh less likely.

BEAVER VALLEY - UNIT 1 B 3/4 4-3h Amendment Chanqe No. -94&1-029 l

REACTOR COOLANT SYSTEM ProvidedforInformation Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS (Continued) ab. If any pressure boundary LEAKAGE exists or primary to secondary LEAKAGE is not within limit, or if unidentified or identified LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 34 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 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)

SR 4.4.6.2.a An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor the containment atmosphere radioactivity and the containment sump level. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring of the leakage detection system is sufficient to provide an early warning of increased RCS LEAKAGE. These leakage detection systems are specified in LCO 3.4.6.1. "Leakage Detection Instrumentation."

Note (1) states that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance is recquired only on leakage detection instrumentation required by LCO 3.4.6.1. This Note allows the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring to be suspended on leakage detection instrumentation which is inoperable or not required to be operable per LCO 3.4.6.1.

SR 4.4.6.2.b 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. Primary t- seeendary LEAKACE is alAd measured by perOfFrmane f an RGS water inventery balanee in eenjuneti.n with ef fluent monitering within the se*endary steam and fe1dwater oyztomo

M=-

ProvidedforInformation Only.

The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure.

Therefore, this OR is net required to be perfermed in MODES 3 and 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation near operating pressure have been established-.The SR is modified by two notes. Note 2 states that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable Plant conditions are established.

BEAVER VALLEY - UNIT 1 B 3/4 4-3i AmendmentfChange No. 1~841-029

Providedfor Information Only.

REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

SURVEILLANCE REQUIREMENTS (SR) (Continued)

. t SR) A A4. *2 (Gnt Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and a Note requires the Surveillance to be met when steady state is established.

For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. Note (2) states that this SR is required to be performed during steady state operation.

An early warning of pressure beundary LEAKACE er unidentified LEAKACE is provided by the systems that monitor the containment atmesphere radieaetivity and the eentainment sump level. The 12 heur monitoring of th iagOel ortetien systom is suffieient te previde an early warning ef increased RCC LEAKAGE. These leakage detectien systemo are zpecified in LCO 2.4.6.1, "Leakage Deteetien I-natrumentatien."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. Neot (1) states that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance is required only on leakage detcetion instrumentation required by LCO 3.4.C.1. This Note allows the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring to be suspended on lcakagc detcetion instrumentation which is inop-rabl-or not required to be operable per LCO 3.4.6.1. Note (2) states that this SR is required to be performcd during steady state operation.

Note 3 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.

SR 4.4.6.2.c This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.5, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature (250C) as described in EPRI.

"Pressurized Water Reactor Primary-to-Secondary Leak Guidelines". 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. For RCS primary to secondary

Providedfor Information Only.

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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with EPRI.

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

3/4.4.6.3 PRESSURE ISOLATION VALVE LEAKAGE The leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure. It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valve is identified LEAKAGE and will be considered as a portion of the allowed limit.

BEAVER VALLEY - UNIT 1 B 3/4 4-3j AmendmentgChanqe No. 1841-029

Attachment B-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Bases Changes License Amendment Request No. 196 The following is a list of the affected pages:

B-II B 3/4 4-2 B 3/4 4-3 B 3/4 4-3a B 3/4 4-3b B 3/4 4-4d*

B 3/4 44e*

B 3/4 4-4f' B 3/44-4g B 3/4 4-4h B 3/4 4-4i B 3/4 4-4j

Provided for readability only

Provided for Information Only.

TECHNICAL SPECIFICATION BASES INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION . ................................. B 3/4 4-1 3/4.4.3 SAFETY VALVES ......... ....................... B 3/4 4-2 3/4.4.4 PRESSURIZER . ................................. B 3/4 4-2 3/4.4.5 STEAM GENERATOR& (SG) Tube Integrity ......... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .... ........... B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY ....... ..................... B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ..... ............. B 3/4 4-6 3/4.4.11 REACTOR COOLANT SYSTEM RELIEF VALVES .... .....B 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS .......... B 3/4 5-1 3/4.5.2 AND 3/4.5.3 ECCS SUBSYSTEMS .......... B 3/4 5-la 3/4.5.4 SEAL INJECTION FLOW .......... B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT .......... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ......... B 3/4 6-10 3/4.6.3 CONTAINMENT ISOLATION VALVES ..... ..... B 3/4 6-12 BEAVER VALLEY - UNIT 2 B-II Change No. 2-Q,24031

REACTOR COOLANT SYSTEM Provided for Information Only.

BASES 3/4.4.2 (This Specification number is not used.)

3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 345,000 lbs. per hour of saturated steam at the valve set point.

During shutdown conditions (MODE 4 with any RCS cold leg temperature below the enable temperature specified in 3.4.9.3) RCS overpressure protection is provided by the Overpressure Protection Systems addressed in Specification 3.4.9.3.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

Safety valves similar to the pressurizer code safety valves were tested under an Electric Power Research Institute (EPRI) program to determine if the valves would operate stably under feedwater line break accident conditions. The test results indicated the need for inspection and maintenance of the safety valves to determine the potential damage that may have occurred after a safety valve has lifted and either discharged the loop seal or discharged water through the valve. Additional action statements require safety valve inspection to determine the extent of the corrective actions required to ensure the valves will be capable of performing their intended function in the future.

3/4.4.4 PRESSURIZER The requirement that 150 kw of pressurizer heaters and their associated controls and emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.

MA 4 A CIMG'7aR One OPERABLE steam generater in a non isolated reacter coolant loop provides sufficient heat removal capability to remove decay heat after a reactor shutdown. The requiremont for two OPERABLE steam generaters, cembined witch otho,,r- reoquirements of the Limiting Cenditiens fer Operatien ensures adequate BEAVER VALLEY - UNIT 2 B 3/4 4-2 Change No. 2-4-21-031 I

REACTOR COOLANT SYSTEM ProvidedforInformation Only.

BASES

-3/4.4.' STEAMS GENBRATORG (Gentinued) decay heat removal capabilitics fcr RCS temperaturcs greater than soor' if onc steam gencrator becomes inoperable due to single failure considerations. Below 3500 F, decay hcat is removed by the RIR eystem.

The Aurvoillance Rquiro4ntos fr in:pctien of the steam gonorator tubes ensure that the otructural integrity of this portion of the RCS will be maintained. The program for inscrvicc inspection of steam genorator tubes is based en a modification of Regulatory Cuidc 1.83, Revision 1. Inscrvicc inspection of steam genorator tubing is essential in erder to mainHtain ourveillance of the eenditiens ef the tubes in the event that there is evidenee ef mocehanieal damage or progr__tiv- d4gradatien due te design, m.anufaeturing -ArrAr4, o r-insArviee eonditinos that lead to err n. iA__rvieA

n. insp__tien of oteam gelnerater tubing a]oo prvdes a means of eharaeterizing thoe nature and eause of any tube degradation se that eerreetive measures ean be taleen.

The plant is expected to be ecperated in a mnanner such that the seeondary eeelant will be maintained within these paramoeter limito found te result in negligible eerresien of the steam gonorator tubes.

if the seeondary eoolant chefmistry is not maintained within theze parameter- limits, leealized eor-resion mnay likely result in stress c-r--r_ vn_racking. The eEtent of -racking during plant eperati-on-would be limited by the limitatien of steam gonorator t~ube-leakagoe between the Primary Coolant Cyotom and the Seeondary Coolant SyAtem (primflary to seendar- L _AKGB 150 gallon. per day per steam generater). A3al er-acko having a primaary to oeeondary LBAKAGCE le&&

than this limit durring eperatien will have an adequate nar-gin of safoty. to withstand the leads impesed duri'ng normnal eperatien and-by pootulated accidents. eperating plants have deomonotrated thait-primary to o ocondary LEAKAGCE of 150 galleno per day per oteam gcnorator can readily bc deteeted. LEAKAGE in excooo of this limait-will1 require plant shutdewn and an unscheduled inspeetien, during which the beaking tibes :will be located and plugged or repaired by sleeving. The technieal bases for oleeving arc described in the approved vendor reperto listed in- Curveillance Requireoment 4.4..4.a.9.

Wastage type defeets are unli:kely with the all volatile treatfent-(AYT-) of seeondary coo-lant. Itowever, even if a defect of sifmilar-typo should develep in serviee, -it will be found during ocheduled-inoervico steam gcncrator tube examinatiens. Plugging or repair will be required of all tubes with imperfections oe~cooding the plugging or, repair limit. Degraded oteam gonorator tubes mnay be repaired by the inotallation of oleeves which span the degraded tube ocetien. A steam gonorator tube with a sleeve installed mootoe the otructural BEAVER VALLEY - UNIT 2 B 3/4 4-3 AfnendmentChanQe No. G02--031 1

REACTOR COOLANT SYSTEM I Provided for Information Only.

BASES

/14.41.5 STEAM GENIERATORS (Gentinued roquiroments of tubes which arc not degraded, therefore, the sleove is considered a part of thc tube. The surveillance requiremcntz identify thoec zleeving methodologiez approved for usc. If an installcd sleevc is found to havc through wall pcnctration greater than or equal to thc plugging limit, the tube must be plugged. The plugging limit for the sleeve is derived from R. C. 1.121 analysis which utilizes a 20 perecnt allowance for eddy currcnt uncertainty in determining the depth of tube wall penetration and additional degradation growth. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detcet degradation: that has pcnetrated 20 perecnt of the original tube wall thickenesc.

Thc voltage based repair limito of thecs ourveillance requireemnto (SR) implement the guidance in CL 95 05 and arc applicable only to Westinghousc decigned steam generators (SCc) with outsidd diameter stress corrosion eraeking (GDSCCC) !ocated at the tube to tube suppert-platc interscetions. Thc guidance in CL 95 05 wvill not bc applied to the tube to flow distribution baffle plate interseetieno. The voltage bascd repair limits arc not applicablc to other forms of £C tube o gradatien nor are they applicabl_ to _DSG that _ecur_ at; ether loation within the SG. Additionally, thp rVpair _ritria apply enly to indieations where the degradation mechanism ic-dcminantly axial GDCCC wi~th no NDE detectable cr-aeks extending outside the thickness of the suppert plate. Refer to CL 95 05 f4r additional deseriptien of th deogradation morphelegyt.

Implemnetation of thesc CRs requircs a derivation of the voltage

.ctructural limit from tnc Dur:c ver-uc votagc ompirica* zorreaeiafn and then the subscquent derivation of the voltage repair limit from thc otructural limit (which is then implemented by thia surveillance).

The voltage otructural imni t i the voltage from the burst, pressure/bebbin voltage correlation, at the 95 pereent predietion.

interval eurve rodueed to account for the lewer 95/95 pereent toleranco bound for tubing mnateri~al properties at 6SO0 P (i.e., the 95 pereent LTL3 curve). The veltage structural limit ffust be adjusted dewnward to account for potential degradatien growth during a operating interval and to aceount for !ME uneertaint-y. The upper-voltage repair limit,- s dt-rmined from the structural voltage i_,

limit by applying thol folwn qatien.

BEAVER VALLEY - UNIT 2 B 3/4 4-3a Amendinen--'n q No. 10-12-031 I

REACTOR COOLANT SYSTEM Providedfor Information Only.

BASES 3/1.1.5 ST-BAD! GBNrUAT-ORS (Gentinued where-V. reprcsents the allowance for degradation growth between of crror in the mFacurement of the bobbin coil voltage. Further diccussion of the assumptions necessary to determine the voltage repair limit arc discussed in CL 95 05.

Safety analyses were perfer-med pursuant te Cenerdic lotter 95 GE; to doetermfine the mawtinvum ?4GLB indueed primary te ceeondary leak rate that eeuld occur witheut effcite docec ei~eeeding a small fraetion oef 10 CER 100 (eeneurrent iedine spilee), 10 CFR 100 (pre aeeident iedine cpike) , and witheut contrel reoom desec exeeeding CDC 1:9. The eur-r-nt-value of the fnaxifmum MC9LD induecd leake rate and a cummifary of- the analyses are previded in Seetien 15.1.5 of the UFCA.R.

T-he mid cyele equatioen in SR 4.45.4.a.969d cheuld enly be used during unplanned inspeetienc in which eddy curront Elata icaurd for indieatiens at the tube cuppcrt plates.

SR 4.45.5 implements several reper-ting regirmnt rcmmended by CL 95 05 for cituatien which the NRC w ntc to be notified prior- to returni:ng the Scc to cervicee. For the purpococ of this reperting r31ir4 ont, leakage and conditional burst probability can be calculatoed based on the as found voltage distributien rather than the projected end of cyele (BGCC) voltage distribution (refer to CL 95 056 for mere informnation) when it is not practical to complete thece ealeulations using the prejected EGC voltage distributiens prior to returning the SCs to corvice. Nete that if leakage and conditional burst probability were calculated using the measured EOC veltage dictributien for the purpocec of addressing the GL coctien 6.a.1 and 6.a.3 reperting cri~teria, then the resultc of the projeeted EOGC veltage dictribution sheuld be provided per the CL cectien 6.b (e)-

criteri~ar Whenevr the re ultc of any steam goenratr tubing incervico inspeetien fall into Categery C 3, thece results will be reperted toe the Comnmiccien purcuant to Speeifieatien 6.6 prior to recumptien of-plant eperatien. Guch eacec will be concidered by the omcino a ease by eace basis and mnay recult in a requirement for analycic, laboratory eoEaminatienc, tects, additional eddy current inspeetioen, and revicien cf the Technical CSpeifiatienc, if nco.ssary.

3/4.4.5 Steam Generator (SG) Tube Integrity BACKGROUND Steam generator tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions.

Steam generator tubes are an integral part of the reactor coolant

Provided for Information Only.

pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system.

This Specification addresses only the RCPB integrity function of the SG. The SG heat removal- function is addressed by "Reactor Coolant Loop" LCOs 3.4.1.1 (MODES 1 and 2), 3.4.1.2 (MODE 3), and 3.4.1.3 (MODES 4 and 5).

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Depending upon materials and design, steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 6.19, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.19, tube integrity is maintained when the SG performance criteria are met.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 6.19. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used-to mLeet the SG performance criteria are defined by NEI 97-06, "Steam Generator Program Guidelines".

APPLICABLE SAFETY ANALYSES The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary SG tube LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.6.2.c, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes that following reactor trip the contaminated secondary fluid is released to the atmosphere via safety valves. Environmental releases before reactor trip are discharged to the main condenser.

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.) In support of voltage based repair criteria, analyses were performed pursuant to Generic Letter 95-05

ProvidedforInformation Only.

to determine the maximum main steam line break (MSLB) induced primary to secondary leak rate that could occur without offsite doses exceeding a small fraction of 10 CFR 100 (concurrent iodine spike),

10 CFR 100 (pre-accident iodine spike), and without control room doses exceeding GDC-19. The established maximum is provided to evaluate leakage projections in lieu of recalculating the radiological consequence for a MSLB event when projections change.

In the MSLB analysis. the steam discharge to the atmosphere includes primary to secondary SG tube LEAKAGE from all SGs which increases to 2.8 gallon per minute as a result of accident induced conditions. The Control Rod Eiection. the Locked Rotor and the Loss of AC Power accident analyses assume that the post-accident primary to secondary SG tube leakage will remain within the operational leakage limit of 150 qpd per SG as specified in LCO 3.4.6.2.c. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.8. "RCS Specific Activity." limits. Pre-accident and concurrent iodine spikes are assumed in accordance with applicable regulatory guidance. 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 events are within the limits of GDC 19. 10 CFR 100. 10 CFR 50.67 or the NRC approved licensing basis (e.g.. a small fraction of these limits), as applicable.

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 that all SG tubes that satisfy the repair criteria be plugged or repaired in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is repaired or removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged or repaired, the tube may still retain tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube. including the tube wall and any repairs made to it, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.19. "Steam Generator Program." and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity.

accident induced leakage, and operational LEAKAGE. Failure to meet anv 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 integritv of the SG tubes under

Providedfor Information Only.

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 secondarv 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 and Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes", August 1976.

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR. is within the accident analysis assumptions. Except for SGTR and MSLB, accident analysis assumes that accident induced leakage does not result in a tube leakage per SG in excess of the operational leakage limit provided in LCO 3.4.6.2.c of 150 qpd per SG. As explained in "Applicable Safety Analyses", the MSLB analysis established the maximum acceptable MSLB induced primary to secondary leak rate. Total leakage may increase to 2.8 qpm during a MSLB event. Accident induced leakage does not exceed 150 cpm per SG.

except for specific types of degradation at specific locations where the NRC has approved greater accident induced leakage. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.6.2. "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY ProvidedforInformation Only.

Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1. 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1. 2, 3. and 4. In MODES 5 and 6. primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the actions may be entered independently for each SG tube. This is acceptable because the recruired actions provide appropriate compensatory actions for each affected SG tube. Complying with the reguired actions may allow for continued operation. and subsequently affected SG tubes are governed by subsequent condition entry and application of associated recruired actions.

a. ACTION a applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged or repaired in accordance with the Steam Generator Program as required by SR 4.4.5.1. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged or repaired has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Action b applies.

A completion time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, ACTION a allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.

However, the affected tube(s) must be plugged or repaired Prior to entering MODE 4 following the next refueling outage or SG inspection. This completion time is acceptable since operation until the next inspection is supported by the operational assessment.

Provided for Information Only.

b. If the required actions and associated completion times of ACTION a are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The allowed completion times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REOUIREMENTS SR 4.4.5.1 During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, "Steam Generator Program Guidelines" and its referenced EPRI Guidelines establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program in conjunction with the degradation assessment determines the scone of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program and the degradation assessment also specify the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 4.4.5.1. The Frequency is determined by the operational assessment and other limits in EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines". The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.19 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 4.4.5.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is repaired or removed from service by plugging. The tube repair criteria delineated in Specification 6.19 are intended to ensure that tubes accepted for continued service

I..

II ProvidedforInformation Only. I a

satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s).

NEI 97-06 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

Steam generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program.

The Frequency of "prior to entering MODE 4 following a SG inspection" ensures that SR 4.4.5.2 has been completed and all tubes meeting the repair criteria are plugged or repaired prior to subjecting the SG tubes to significant primary to secondary pressure differential.

DEAVER VALLEY - UNIT 2 B 3/4 4-3b Amedeffint-hA-nge No. 1:012-031 l

REACTOR COOLANT SYSTEM R Providedfor Readability Only.

BASES 3/4.4.6.1 LEAKAGE DETECTION INSTRUMENTATION (Continued)

SURVEILLANCE REOUIREMENTS (SR)

SR 4.4.6.1.a SR 4.4.6.1.a requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

SR 4.4.6.1.a requires the performance of a CHANNEL FUNCTIONAL TEST on the required containment atmosphere radioactivity monitor.

The test ensures that the monitor can perform its function in the desired manner. The test verifies the alarm setpoint and relative accuracy of the instrument string. The Frequency of 31 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

SR 4.4.6.1.a also requires the performance of a CHANNEL CALIBRATION on the required containment atmosphere radioactivity monitor. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 18 months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this Frequency is acceptable.

SR 4.4.6.1.b SR 4.4.6.1.b requires the performance of a CHANNEL CALIBRATION on the required containment sump monitor. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 18 months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this Frequency is acceptable.

3/4.4.6.2 OPERATIONAL LEAKAGE BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

BEAVER VALLEY - UNIT 2 B 3/4 4-4d Amendment No. 64

REACTOR COOLANT SYSTEM R Provided for Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

BACKGROUND (Continued)

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30, requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100 percent leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB)- from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

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 an event resulting in steam discharge to the atmosphere assumes a 1 gpm 150 god per steam generator (450 gpd total) primary-to-secondary LEAKAGE as the initial condition. An exception to the primary-to-secondary LEAKAGE is described below for the main steamline break (MSLB) analyzed in support of voltage-based steam generator tube repair criteria.

BEAVER VALLEY - UNIT 2 B 3/4 4-4e Amendment-Chfic No. 2-z03

-0 REACTOR COOLANT SYSTEM R ProvidedforReadability Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

APPLICABLE SAFETY ANALYSES (Continued)

Primary-to-secondary LEAKAGE is a factor in the dose releases outside containment resulting from a MSLB 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 MSLB is more limiting for site radiation releases. The primary-to-secondary LEAKAGE assumed in the safety analysis for the MSLB accident is described in UFSAR Section 15.1.5. The radiological consequences of a MSLB outside of containment was reanalyzed in support of the tube support plate voltage-based repair criteria stated in SR 4.4.5.4.a.10. For this analysis, the thyroid dose was maximized at 10% of the 10 CFR Part 100 guideline of 300 rem for the co-incident iodine spike case. RCS leakage was based on projection rather than on technical specification leakage limits. The analysis indicated that offsite doses would remain within regulatory criteria with the assumed primary-to-secondary leakage (described in UFSAR Section 15.1.5) should steam generator tubes fail due to the depressurization associated with a MSLB.

A similar analysis was performed using a control room thyroid dose of 30 rem as the criterion. The control room was assumed to be manually isolated and pressurized at T=30 minutes for a period of one hour, at which time filtered emergency intake would be automatically started.

The control room would be purged with fresh air at T=8 hours following release cessation. The analysis indicated that control room doses would remain within regulatory criteria with the assumed primary-to-secondary leakage (described in UFSAR Sectionc;-15.1.5) should steam generator tubes fail due to the depressurization associated with a MSLB.

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB.

LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Should pressure boundary LEAKAGE occur through a BEAVER VALLEY - UNIT 2 B 3/4 4-4f Amendment No. 101

REACTOR COOLANT SYSTEM Providedfor Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

LCO (Continued) component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Primary- to- Secondary LEAKAGE through Any One SG Operating cepericnce at PWR plants has shown that sudden inareases in leak rate are eften preeursers te larger tube falrc_ 4ana n an saprtn A. _AC limi7t af\' :.^

L_4E4 15 gp p .cr zita g ancrat wil miis th patantia far a lrg ,

LEAIC -s I._~ AvnI . -AI _ p--.-

al ar.

A Thic Y_ S.t 4 :crSa-sti

- -- -- Y-LEP_

-- CE lii i_ mVr ratrctv tha t_ _---tin L-*_ .- _ _l limitF-- X_

in standardiczd tcchnncal specellcatlens. inis pra' additional margin to accommodatc a tubc flaw which i grow at a grcatcr than expeeted ratc or unciepectedly c:

outsidc the thickncss of the tubc support platc. -This reduced LEAKACE limit, in canjunction with a lcakc monitoring program, providcs additianal assuranac that -atew wrecuTs rLEAXAP-P will b 1=wn te andi thc1p nlant:su in a timcly manncr.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. The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states. "The RCS operational primary to secondarv leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in coniunction with the implementation of the Steam Generator Program is an effective measure for minimizing the fremuency of steam generator tube ruptures.

BEAVER VALLEY - UNIT 2 B 3/4 4-4g AmendmentfChanqe No. 1-0-12-031 I

REACTOR COOLANT SYSTEM Providedfor Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

LCO (Continued)

d. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.6.2, "RCS Pressure Isolation Valve (PIV)," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

BEAVER VALLEY - UNIT 2 B 3/4 4-4h AmcndmentChange No. 1-0-12-031 I

REACTOR COOLANT SYSTEM R Provided for Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS ba. Unidentified LEAKAGE-- or identified LEAKAGE, or primary to scoendary LEAKACE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

It the unidentiticed L IAKACE, idientified LEAKACE, or primary to secondary LEAKACE eannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKGCE and its petential eensequences. The reaeter must be brought te MODE a within 6 heurs and MODE S within 36 heurs. This actien reduces the LEAAG~.E~.-

The allewed Cempletien Times are reasenable, based en eperating emperrienee, to roeach the -required plant eenditiens from full pewor cenditiene in an orderly manner-and without challenging plant systems. in MODE 5, the pressure stresses aeting on the RC-P are much !ewer, and further deterieratien is much less likely.

ab. If any pressure boundary LEAKAGE exists- or primary to secondary LEAKAGE is not within limit, or if unidentified or identified LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 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 broucht to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 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.

BEAVER VALLEY - UNIT 2 B 3/4 4-4i AmendmentChanqe No. .1--2-031 1

REACTOR COOLANT SYST'EME-R Provided for Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

SURVEILLANCE REQUIREMENTS (SR)

SR 4.4.6.2.a I An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is Provided by the systems that monitor the containment atmosphere radioactivity and the containment sump level. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring of the leakage detection system is sufficient to provide an early warning of increased RCS LEAKAGE. These leakage detection systems are specified in LCO 3.4.6.1, "Leakage Detection Instrumentation."

Note (1) states that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance is required only on leakage detection instrumentation required by LCO 3.4.6.1. This Note allows the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring to be suspended on leakage detection instrumentation which is inoperable or not required to be operable per LCO 3.4.6.1.

SR 4.4.6.2.b 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. Primary to secondary LEAKCAE is also measured by p1rfermanee ef an RC' water invnt ry balance in oenjuneti.n with ef fluent monitering within the seeendary steam and feedwater zyetems.

The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure.

Therefore, this SR is not required to be performed in MODES a and 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation near operating pressure have been established.The SR is modified by two notes. Note 2 states that this SR is not recquired to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and a Note requires the Surveillance to be met when steady state is established.

For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. Note (2) states that this SR is required to be performed during steady state operation.

An early warning of pressure boundary LEAYRCE or unidentified LEAKGCE is provided by the systems that monitor the containment atmosphere radioactivity and the containment sump level. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring of the leakage detettion zystem is sufficient to provide an early

U ProvidedforInformation Only. I warning of increased Rcs LEAKACE. These leakage detoetion system&

arc specified in LC 3.1.4..1, "L3akage Detoetion Instrumentation." I The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. Note (1) states that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance i G required only on lealeage deotetion instrumentation required by LCO 3.4.6.1. This Neot allows the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring to be suspended on lakeaag detettion instrumentation which is inoperable or not required to be operable per LCO 3.4.6.1. Note (2) states that this SR is required to be performed during steady state operation.

Note 3 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.

SR 4.4.6.2.c This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primarv to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met. compliance with LCO 3.4.5, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature (250C) as described in EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines". 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 secondarv 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 Freguency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with EPRI, "Pressurized Water Reactor PrimarV-to-Secondary Leak Guidelines".

BEAVER VALLEY - UNIT 2 B 3/4 4-4j AmendmentChanqe No. 102-031 l

Attachment C Beaver Valley Power Station, Unit Nos. I and 2 License Amendment Request Nos. 324 and 196 Comparison to TSTF-449

Attachment C LARs 324 and 196 Comparison to TSTF-449 TSTF Specification TSTF Change Description Corresponding BVPS Change Description Comments Proposed BVPS Specification 1.I - Definitions Revised definition of LEAKAGE 1.14.a.3 and 1.14.c Definition would be revised as proposed by the TSTF.

LCO 3.4.4, 3.4.5, 3.4.6, Deleted reference to Steam Generator N/A No change. BVPS SG tube surveillance and 3.4.7 Bases - RCS Tube Surveillance Program. requirements are not yet located Loops in the administrative section of the TS. Therefore there are no existing references to it that must be revised. (Refer to comments

._ for 3.4.20.)

LCO 3.4.13 - RCS Deleted limit of I gpm primary to LCO 3.4.6.2 No change. Current LCO does Current LCO overall is already Operational Leakage secondary leakage through all SGs not include this limit. consistent with the revised LCO Revised limit of [500] gpd primary to No Change. Current LCO proposed by the TSTF.

secondary leakage through any one SG already contains a 150 gpd limit However, minor editorial to 150 gpd as proposed in the TSTF. changes would be made.

LCO 3.4.13 Bases Deleted basis for I gpm leakage limit LCO 3.4.6.2 Bases No change. The current BVPS through all SGs LCO does not specify this limit.

Replaced basis for primary to secondary Replaced as proposed by TSTF leakage through any one SG Actions 3.4.13.A and B Insert "operational" before "LEAKAGE" Action 3.4.6.2.b and a Revised as proposed by TSTF and The current BVPS TS actions throughout 3/4.6.2 have the same meaning as the Exclude primary to secondary leakage Revised as proposed by TSTF standard, but differ in format and from conditions that allow 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months wording. Therefore, in addition restoration in Action A to incorporating technical Add primary to secondary leakage to Revised as proposed by TSTF changes proposed by the TSTF, conditions that require plant shutdown in these actions would be Action B rearranged and edited to use wording similar to the TSTF wording.

Page I of 5

Attachment C LARs 324 and 196 Comparison to TSTF-449 TSTF Specification TSTF Change Description Corresponding BVPS Change Description Comments Proposed IBVPS Specification Action 3.4.13.A and B Reflect exclusion of primary to Action 3.4.6.2.b and a Revised as proposed by TSTF Refer to comments for Actions Bases secondary leakage from conditions that bases 3.4.13.A and B above.

allow 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restoration in Action A Reflect addition of primary to secondary Revised as proposed by TSTF leakage to conditions that require plant shutdown in Action B SR 3.4.13.1 Inserted Note 2 SR 4.4.6.2.b Inserted new Note 3 with wording proposed by TSTF Capitalized "leakage" No change. BVPS TS do not have lower case "leakage" N/A Revised Note 2 to be consistent with TSTF wording. This includes removing "in MODE 3 or 4" from the note.

SR 3.4.13.1 Bases Remove primary to secondary leakage SR 4.4.6.2 Bases Revised as proposed by the The existing bases have a high detection from SR basis TSTF. degree of similarity to the Added discussion of new Note 2 Added discussion of new Note 3 standard. However, due to as proposed by TSTF. differences that exist, some of N/A Renumbered as 4.4.6.2.b the revisions proposed by the N/A TSTF would be incorporated with slight modifications that maintain the intended meaning.

SR 3.4.13.2 Replaced the SR with requirement to SR 4.4.6.2.c Inserted new SR with same verify primary to secondary leakage is requirement as proposed by TSTF less or equal to 150 gpd Revised frequency to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months New SR contains this frequency Inserted a Note New SR contains this note SR 3.4.13.2 Bases Revised to reflect new 150 gpd SR, SR 4.4.6.2.c - bases Basis for new SR 4.4.6.2.c would frequency and note. be inserted as proposed by TSTF.

Page 2 of 5

Attachment C LARs 324 and 196 Comparison to TSTF-449 TSTF Specification TSTF Change Description Corresponding BVPS Change Description Comments Proposed BVPS Specification 3.4.13 Bases - Applicable Revised to reflect operational leakage 3/4.4.6.2 Bases - No change. The current BVPS LCO 3.4.6.2 Safety Analyses specification changes Applicable Safety is already consistent with the Analyses LCO proposed by the TSTF. No accident analyses have been changed. Therefore the corresponding bases which contain a plant-specific description of applicable safety analyses do not require revision.*

3.4.13 Bases - References Inserted items in reference listing 3/4.4.6.2 Bases The new references would be incorporate into the bases at the point of reference rather than in a list of references. _

3.4.20 - Steam Generator Inserted a new specification 3/4.4.5 - Steam Replaced current specification Changes proposed by the TSTF Tube Integrity Generators with a new one containing the presume that SG tube same requirements as proposed surveillance requirements have by the TSTF. been relocated from this LCO to the administrative section of the technical specifications. The TSTF then proposes changes to the relocated requirements.

Since BVPS tube surveillance requirements are not yet relocated, proposed changes are intended to reflect both relocation and incorporation of subsequent changes proposed by the TSTF in a single step. The current 3/4.4.5 contains the requirementstbeing relocated.

For BVPS-2, a correction to the stated primary to secondary leak rate assumption used in safety analyses is shown on page B 3/4 4-4e.

This change should have been incorporated with License Amendment 118.

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Attachment C LARs 324 and 196 Comparison to TSTF-449 TSTF Specification TSTF Change Description Corresponding BVPS Change Description Comments Proposed BVPS Specification 3.4.20 Bases Inserted bases corresponding to the new 3/4.4.5 - Bases Replaced current specification Refer to comments for 3.4.20 specification bases with new ones proposed by above.

the TSTF.

The "Applicable Safety Analyses" section of these bases reflects plant-specific information that differs from the model bases.

The "LCO" section of these bases reflects accident induced leakage rate values that are consistent with plant-specific accident analyses instead of the generic I gpm value contained in the model TS.

5.5.9 - Steam Generator Replaced steam generator tube 6.19 - Steam Generator Inserted a new administrative BVPS SG tube surveillance Tube Surveillance surveillance program requirements with Program specification requiring a steam requirements are not yet located Program a requirement to maintain a steam generator program. in the administrative section of generator program. The program would the TS. Therefore 6.19 would be contain the surveillance requirements. a new specification. (Refer to comments for 3.4.20)

TS 6.19.6.2 reflects accident induced leakage rate values that are consistent with plant-specific accident analyses instead of the generic I gpm value contained in the model TS.

Existing BVPS-2 repair criteria for tube sleeves is included in TS 6.1.9.c.

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Attachment C LARs 324 and 196 Comparison to TSTF-449 TSTF Specification TSTF Change Description Corresponding BVPS Change Description Comments Proposed BVPS Specification l 5.6.9 - Steam Generator Replaced reporting requirements with 6.9.7 - Steam Generator Inserted a new administrative BVPS SG tube inspection Tube Inspection Report revised ones. Tube Inspection Report specification for reporting as reporting requirements are not proposed by the TSTF. yet located in the administrative section of the TS. Therefore, 6.9.7 would be a new specification. (Refer to comments for 3.4.20 above.)

Existing reporting requirements associated with BVPS-2 alternate repair criteria have been added to this specification.

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v t e Letter Sequence Request
TAC:MC8861, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request Acceptance Supplement, Supplement Administration Questions 1 March 2006 Answers 25 April 2006 Results Approval Other:
MONTHYEAR2006-04-25 [Table View]

L-05-144, License Amendment Request Nos. 324 and 196 Re Steam Generator Tube Integrity

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1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

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6.0 REGULATORY ANALYSIS

10.0 REFERENCES

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L-05-144, License Amendment Request Nos. 324 and 196 Re Steam Generator Tube Integrity

Text

Subject:

Enclosure:

Subject:

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

10.0 REFERENCES

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

10.0 REFERENCES

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