Text

Supplement to License Amendment Request Nos. 324 and 196, Steam Generator Tube Integrity
	ML062210423
Person / Time
Site:	Beaver Valley
Issue date:	08/03/2006
From:	Lash J; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-06-119, TAC MC8862, TAC MC8861
	Download: ML062210423 (93)
	v • d • e

FENOC FirstEnergy Nuclear OperatingCompany James H. Lash 724-682-5234 Site Vice President Fax: 724-643-8069 August 3, 2006 L-06-119 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 Supplement to License Amendment Request Nos. 324 and 196 Steam Generator Tube Integrity (TAC Nos. MC8861 and MC8862)

By letter dated June 1, 2006 (L-06-088), the FirstEnergy Nuclear Operating Company (FENOC) submitted a supplement to License Amendment Request (LAR) Nos. 324 and 196 that would revise steam generator tube integrity technical specifications for Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2. Subsequently, by E-mail message dated June 19, 2006, the NRC provided several draft questions involving clarity and precision of proposed technical specification wording in the June 1, 2006 FENOC submittal.

These were subsequently discussed in a teleconference on July 13, 2006. A follow-up E-mail message was received on July 18, 2006. Attachment A provides responses to the NRC staff's draft questions in the June 19, 2006 E-mail message.

Attachments B-1 and B-2 are proposed BVPS-1 and BVPS-2 Technical Specification (TS) changes. Attachments C-1 and C-2 are proposed BVPS-1 and BVPS-2 TS Bases changes. Changes in these attachments include refinements that incorporate responses to NRC draft questions in the June 19, 2006 E-mail message, resolution of a comment received in the July 18, 2006 E-mail message, and minor editorial enhancements and corrections that do not affect the intended meaning. The proposed TS Bases changes are provided for information only. These attachments supersede the corresponding attachments contained in the initial LAR submittal and the June 1, 2006 supplement.

FENOC has determined that the revisions proposed by this supplement do not affect the original evaluation of proposed changes or No Significant Hazards Consideration Determination provided in the November 7, 2005 submittal.

No new regulatory commitments are contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Gregory A. Dunn, Manager, FENOC Fleet Licensing, at (330) 315-7243. P

Beaver Valley Power Station, Unit Nos. 1 and 2 Supplement to License Amendment Request Nos. 324 and 196 L-06-119 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on August -3 ,2005.

Sincerely,

(,A James H. Lash Attachments:

A Responses to Draft Questions Regarding June 1, 2006 LAR Supplement B-1 Proposed BVPS-1 Technical Specification Changes B-2 Proposed BVPS-2 Technical Specification Changes C-1 Proposed BVPS-1 Technical Specification Bases Changes C-2 Proposed BVPS-2 Technical Specification Bases 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)

Attachment A to L-06-119 FENOC Response to Draft Questions Regarding June 1, 2006 Supplement to License Amendment Request Nos. 324 and 196 Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2)

Steam Generator (SG) Tube Integrity Technical Specification (TS)

1. Please discuss your plans to correct the typographical error in LCO 3.4.6.2 Action a for Unit 2. This LCO should read as follows:

"With any Reactor Coolant System operational Leakage not within limits for reasons other than..."

Response

Typographical error "then" has been corrected to "than" in the proposed TS.

2. Given that Technical Specification Section (TS) 6.19.c.4 does not provide safety factor requirements, please discuss your plans to modify 6.19.b.1 to further clarify that all flaws except the flaws addressed in 6.19.c.4 will include a safety factor of 3.0 against burst under normal power operation. For example, "This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 6.19.c.4,..."

Response

Based on a July 13, 2006 teleconference with NRC staff, it is understood that proposed wording related to the 1.4 safety factor could be misconstrued to mean that proposed BVPS-2 TS 6.19.c.4 contains an alternative safety factor requirement. Therefore, proposed TS 6.19.b.1 has been revised to clarify that the reference to TS 6.19.c.4 pertains to flaws rather than to the safety factor by stating, "...except for flaws addressed through application of..." instead of"...except as permitted through application of..."

In addition, further clarification needs to be made regarding which indications are limited to a probability of burst under postulated main steam line break conditions less than 1X10 2 . As currently written, once the alternate repair criteria in 6.19.c.4 is implemented, the probability of burst for all indications (even those that are not axially oriented outside diameter stress corrosion cracking at tube support plate locations) is limited to lX10"2. For example, "When alternate repair criteria discussed in Specification 6.19.c.4 are applied to axially oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than 1X1O 2."

Attachment A to L-06-119 Page 2 of 4

Response

Proposed TS 6.19.b.1 has been more precisely worded as suggested above.

3. TS Section 6.19.c.2 and TS Section 6.19.c.3 appear to contradict each other. TS Section 6.19.c.2 provides a specific plugging limit for a sleeve, whereas TS Section 6.19.c.3 indicates that a tube will be plugged regardless of flaw depth if there is a flaw in the sleeve at the sleeve-to-tube joint. Please discuss your plans to modify TS 6.19.c.3 to remove reference to flaws in the sleeve portion of the sleeve-to-tube joint.

Alternatively, discuss your plans to modify your TS to reflect current industry practice. Namely remove 6.19.c.2 and modify 6.19.c.3 to indicate that tubes with a flaw in a sleeve or in the original tube wall of the sleeve-to-tube joint shall be plugged.

Response

Proposed BVPS-2 TS 6.19.c.2 and 6.19.c.3 do not contradict each other; however, they do provide two differing criteria that each result in plugging of a tube based on flaws found in a sleeve at the sleeve-to-tube joint. Since only the most restrictive criterion needs to be specified, the scope of proposed TS 6.19.c.2 has been revised to limit application of the percent through-wall repair criterion to the non-joint portion of a sleeve. To avoid possible confusion due to wordiness, proposed TS 6.19.c.3 has been revised to more concisely describe applicability to the joint only.

4. In the Applicable Safety Analysis section associated with the Bases for 3/4.4.5, "Steam Generator (SG) Tube Integrity," it was indicated that the analysis for most design basis accident and transients, other than a SG tube rupture, assume that the SG tubes retain their structural integrity. This section further goes on to indicate that an exception to this assumption that tubes retain structural integrity is applied to the Unit 2 steam line break analysis. The basis for such a statement is not clear.

Implementation of the voltage-based repair criteria limits the likelihood that a tube will burst under steam line break conditions by imposing a limit of 1X10-2 on the probability of burst. As a result of this limit, the accident induced leakage methodology assumes that tubes retain their structural integrity during design basis accidents. Please discuss your plans to modify your Bases to make it consistent with the staff's original approval of the voltage-based repair criteria discussed in Generic Letter 95-05.

Response

Terminology used in the proposed BVPS-2 TS Bases has been revised to avoid inconsistency with the concept of structural integrity described in the staff's approval for the use of voltage based repair criteria.

In addition, the staff notes that you indicated that you would be deleting the technical basis for the voltage-based repair criteria from your TS Bases. The reason for this is not clear. Please discuss your plans for including the technical basis for the voltage-based tube repair criteria in your TS Bases.

Attachment A to L-06-119 Page 3 of 4

Response

Proposed BVPS-2 TS Bases statements regarding voltage-based repair criteria would be removed because TS 3/4.4.5 would no longer contain requirements associated with the criteria. Instead, administrative TS 6.19 would contain requirements for a steam generator program, including program requirements for the application of voltage-based repair criteria. This approach is consistent with the philosophy of the TSTF-449 model.

Administrative technical specifications do not have a corresponding TS Bases section.

In your Bases, you indicated that "accident induced leakage" adds 2.1 gallons per minute (gpm) to the total leakage assumed in the Unit 2 steam line break analysis.

You further indicate that you assume there is 150 gallons per day (gpd)

(approximately 0.1 gpm) operational leakage (from each of the three SGs) which when added to the 2.1 gpm "accident induced leakage" results in a total assumed leakage of 2.4 gpm. Since the Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, indicates 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, it appears that your statements are inconsistent with TSTF-449. Please discuss your plans to modify the terminology in your basis to be consistent with the definition of accident induced leakage in TSTF-449. In addition, discuss your plans to re-incorporate this definition into your Bases (i.e., it was deleted in your most recent submittal). For example, based on the staff's understanding of your accident analysis (as written in your current proposed TS Bases), a statement such as the following would be considered consistent with TSTF-449: "In support of voltage based repair criteria, analyses were performed pursuant to Generic Letter 95-05 to determine the maximum main steam line break primary-to-secondary leak rate that could occur without offsite doses exceeding the limits of Title 10 of the Code of FederalRegulations Part 50.67 as supplemented by Regulatory Guide 1.183 and without control room doses exceeding General Design Criteria-19. This analyses requires the leakage from the faulted SG to be limited to 2.2 gpm and the leakage from the non-faulted SGs to be limited to 150 gpd (approximately 0.2 gpm)." In other words, the accident induced leakage for the faulted SG is limited to 2.2 gpm and the accident induced leakage from each of the two non-faulted SGs is limited to 0.1 gpm (or more precisely to 150 gpd per SG).

Please note that there are several places in the Bases that discuss accident induced leakage. Appropriate modifications should be made to all applicable areas (e.g.,

Page B 3/4 4-4f; Applicable Safety Analysis section associated with the Bases for 3/4.4.5).

Response

The example wording provided above indicates that the NRC staff has correctly interpreted the leakage assumptions described in the proposed BVPS-2 TS Bases.

However, use of the phrase "accident induced" throughout the proposed TS Bases is not always consistent with the concept of "accident induced leakage" as applied in the TSTF-449 model. Therefore, the proposed TS Bases have been revised throughout to avoid

Attachment A to L 119 Page 4 of 4 conflicts with this concept. Deleted wording that previously described the meaning of accident induced leakage rate has been reinstated.

In your Bases under Surveillance Requirement (SR) 4.4.6.2.b, Note 2 is discussed twice. Please discuss your plans to remove this redundancy.

Response

The second statement regarding Note 2 has been determined to be unnecessary. This statement has been removed from both the BVPS-1 and BVPS-2 proposed TS Bases.

Attachment B-i 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 V

XV 1-3'*

1-4 3/44-8 3/4 4-9 3/44-10 3/4 4-10a 3/4 4-10b 3/4 4-1Oc 3/4 4-1Od 3/4 4-10e 3/4 4-13 3/44-14 6-21 6-26 6-27*

6-28*

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 GENERATORS (SG) Tube Intearity....... 3/4 4-8 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 V Amendment No. 24ý&

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. 2-6-6 1

DEFINITIONS F SProvided for only readiability 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

T4 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 stea gencrater 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 (QPTR) 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 1-131 1.19 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The DOSE EQUIVALENT 1-131 is calculated with the following equation:

CI_1 3 1 D.E" = CI_131 + CI- 13 2 + CIl-133 + CI-134 + 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. 2-1-6 1

TL REACTOER GG)GO3A~jT ZyCTEM 3' I" "A ATE'll "', l m ,

LIM'ITING CONDITION FOR GPERPqTION 3.4.5 Each steam ganearater shall be OPERABLE.

APPLICABILITY: MOPES 1, 2, 3 and 4.

With .n. mere steam generators

.r in.p.rabla, r.st.r. the in-p.rabl.

generater(s) te OPERABLE status prior te inereasing T~e"abave-2000 F.-

SURVEILLANCE REQUIREMENTS 4.4.5.1 Steam Canerater Samiale Selectien andl lnspatien E a e1h steam gn*nrat"r shall be determined OT"PEERABLE during shutd by saleeting and inspecting at least the minimutm nunia-r ef steam.

gan.rat4rs speifid in Table 4.4 1.

4.4.5.2 Steam Canarater TPuba amle Seleetien and insveetinn The steam generator tube minim.. sample size, inspa-ti.n result elassificatian, and the car-r-spand-ing actien raqfuired shall be as specifiad in Table 4.4 2. The insarvica inspectien ef steam.

ganareatr tubeas shall be perfermed at th,-fragun.i.s spe.ifi. d in Speeifieatien 4.4.5.3 and the inspacted tubes shall be verified-acceptable per the aecaptanca criteria of Specifieatien 4.4.5.4.

Steam ganarater tubes shall be examined in aeeerdanca with Artie!a ef V ("Eddy currant Examinati"n .f asctinn Tubular Pr-ducts") and Appendix 1V to Sctien xi ("Eddy Cur-rant Examinatien ef Nenfarremagnatie Steam Canerater Hecat Emehangar Tubing") of the applicabla year and addenda ef the ACME Bailar and Pres!§ure Vessel Cada required by 10CFR5O, Sactien 50.55a(g). The tubes salactad fear eaah inserviee inspactien shall ineluda at least 3 pareant ef the tetal number af tbas in all stecaa ganaratars;- the tubes salactad fer these inspactians shall be selectad en a randam basis axcapt.

a. W~here axparianea in similar plants with similar water chamistry indieates critical areas te be inspactad, than at least 50 pareant ef the tubas inspactad shall be fram these eritical areas.

b. The first sample af tubas salectad fer eaah insarvi--a inspactien (subsequent ta the prasarvica inspeetian) af eaeh steam ganaratar shall ineluda.

I. All nanpluggad tubas that praviausly had datactabla u-4 penetri-An ~r-OA-Pr-r P I-Rn -,'H I RnRmr flP7'T~flU~T.TF'. 7 TDIrEPT IL 3ý4 44 8 ~ ~ r'r~r r ~ -

REACTO3R COOLANT SYSEPEMI SURVEILLANCE REQUIREMENTS (Czntinued)

2. Tubes in these areas where oxpoionc has indicated potantial preblams, and

3. A tube inspeatien pursuant to Speeifieatien 4.4.5.4.a.8 shall be perfermeed en eaah zsalacted tuba If any seleeted tube doas net permit the passage o-f the ededy eurrent preba for a tube inspeetien, thi-s shall be reeerded and an adjaeent tudba shall b selacted and subjacted to a tube inspeetien.

e. The tubes seleetadl as the seeond and thirdl samples (irf ragir3 by Table 4.4 2) dutring eaeh inserviee inspeetion.

may be subjeeteed te a partial tube inspeetien providad.:

1. The tubes seleeted for these Samples inelude the tubes freim these areas of the tube sheet array where tuba with imperfactions were previeusly fouind, and

2. The inspeetiens iraluide these partions ef the tubas where irnperfeetiens were previeusly fouind.

3/4 4 9 ~L44gAmeandment No. 2:73

l&

REAC-TOR COO-L~a1q SYSTEMi SURVEILLA32CE REQUIREMENTS (Centinued)

The results ef eaeh samnple inspectien shall be elassificd into ene ef the fellewing three eategcrcs rgategery C 1 Less than 5 percent ^f the t*tal tubes inspeeted are degraded tubes an~d nene e-f the inspeeted tubes are defeetive.

C 2 One er mere tubes, but not mere th.n 1 perccnt ef the tetal tubes inspected are d -f"tiv., . 1 between 5 percent and 10 per..nt Cf the t.tal tubes insp "ted ar degraded tubes.

C- 3 Mere than 10 percant Cf the tetal tu-bes Inspcatea are degraded tb*es Cr mere than 1 perccnt Cf the inspecteed tubes arc dcf..tivc.

Note. in all inspections, previ"usly degraded tubes must exhibit signifi.ant (greater than 10 p.r.ent) further wall penetratiefls tC be ineluded in the abeve perccntage ealetlatiens.

4.4.5.3 inspeetien Free~ueneias The abeve required inser-vicc in.p.*ti.ns Cf steam gen.rattr tubes shall be p erformed at the fellewing freeluen -----

a. The first inservic inspe.ti. n Cf the Mcdt l 54F st.a.

g.n.rat.rs shall be ptrfeaed after 6 Effective Full Pewer me~nths but within 24 ealendar menths Cf initial criticality fellewing steam generater replacament.

Suibseqttent inservicc inspeetiens shall be per-fermd at intervals Cf net less than 12 nor mere than 24 ealendar menths after the previeus inspeetien. if two eenseeutive inspeetiens, net ineluding the preservicc inspectien, result in all inspectien results falling intC the C 1 eategery Cr- if twC censeecutive inspectiens demenstrate that pravieusly Cbsarved degradatien has net centinuced and nC additienal dagradatien has Cacurrad, the inspectien interval maay be extended tC a Maxim Cf Cnee par 40 menthe-.

Neta. inservicc inspectien is net requireed during the staai generater replaccment Cutage.

BEAVER VAL13EY bljjjq3 1: 3/4 4 10 BEAVE ~I l V~LY Ne. 2:7-3 3744~QAmendme~nt

REAC.TO.R COOLANT. SY.TE..

ZURVEILLANCE REQUIREMENTS (Continuad)

b. If the results of the -navce inspeetion of a stea ganerator eendutietd in aeeer-danee with Table 4.4 2 fall inte Categery C 3, the inspeetien frequanay shall be inaroasadl to at least onea per 20 months. The inerease J;_

inspeetien fre~aeney shall apply until the subsequent inspeetizns satisfy the eritaria of speeifieatien 4 .4 .5.3 .a; the interval may then be exten~ded to a mrm of enee per 40 months.

e. Additional, unsehadulad insarviee inspeetiens shall b perfermed en eaah steam ganarater- in aeeor-danee with the first sample inspaatien spaaified in Table 4.4 2 during the shultdown subsequent to any of the followingW I. Primary to seeendary tube leaks (net inaluding leales originating from tuba to tuba sheet welds) in emeess of the limits of Spaaifieatien 3.4.6.2,

2. A seismia eeeturrenaa greater than the Gparating Basis O p a q

6. A. +/-Oss eE eeeiane aaaieane raqui~ring aatuaelen Ce the 4 A mqin c~-.-P~nmlinp' A~o ~i-- I iminmb~-

4.4.5.4 Aaaaptanee Critar-ia_

a. As used in this Spaaifieatien:

1. lmiparfeetion means an. axaaptien te the dimansionS, finish or aontetur Cf a tuba from that required by fabriaation drawings or spaaifiaations. Eddyaurn testing~ indieatizrnz balow 20 pereent ef the nerminalI tuba wall thielmass, if detaatable, may be eensidarad as imparfeetions.

2. Daefradatien means a sar~viee indueed eraeleing, wastage, wear or general aorresien eaaurring o~n either inside or outsida of a tuba.

3. Degraded Tuba means a tuba eentaining imparfeetions greater than or equal to 20 pareent of the nomina-l wall thiekness eausad by degradation.

4. Pareant Dagradatien means the percentage of the tuba wall thieaknass affeeted or removed by dagradation.

BEAVER VALLEY

I' REACTOGR COOLTANTP SYSTEM SURV-EILL3ANC-E REQUIREMEN'TS (Cantinued)

5. Dafeet means an iraperfectiar. of sueh severity that it eaeeee 9 the plugging limit. A tube cantaining a daefact is dafeetive. Any tube whieh dacs net pemi-t-the passage ef the eddy eurrant inspe.tien pr-b-shall be deemed a dafective tube.

6. PUeF.i... Liimit means the imp.rf -ti"ndepth at "r beyend which the tube shall be vramaved frerm sar__c by plugging because it may bacame unsarviccabla re tathe next inspectien. The plugging limit is eefual:

ta the 40 percent af the n.minal tuba wall thick-nass.

7. Ugnsarvieeabla daseribes the eenditian ef a tuba if it leaks er caentains" aq dafact large aneugh ta affact its structural integrity in the avant ef an Oparcating Basis Earthefakae, a less af eaclant accidant, ara staamiline ar feedwater line break as spacified in 4.4.5.3.e, ab*va.

8. Tube insvectian means an inspactian af the stea ganaratar tuba fram the peint af entry (hat lag side) camplataly araund the U band ta the tap suppart a the eeca lag.

b. The steam ganaratar shall be determined GPERALE afta eempleting the carraspanding actians (plug all tubas emeeeding the plugging limfit) required by Table 4.4 2.

4.4.6.5 faat

a. Within 15 days fallawing the campletian af eachinrvc inspactian af steam ganaratar tubas, the ntffbr- af tubas plugged in each steam ganaratar shall be submitted in a Special Repart in accordance with 10 CFR 50.4.

b. The w.mplate results f the stea.,mgenerator tuba inr-.v.

inspactian shall be submitted in a Spacial flapart in caeerdanee with 10 CPR 50.4 within 12 months fellawing the eempletien af the inspectian. This Spacial Rapart shall:

4-nelude:e I. Number and extent af tubas inspactad.

2. Lacatian and parcant af wall thicknass penatratian far eaah indicatian af an imparfactian.

"1 p $ * * -- P * "1

.4 rl T R ;q i=r, r- :1 1-.; 4-- :1rv R: Ri t: R4;I TT'" -nI Amendmeant Na. 273

I I 1~ :1 (Nil 1!I-:M :r~r~: U ~flV1 I

c. Results ef steamn gcrerater tube inspeetizr.s whieh fall inte Categery G 3 shall be reperted te the G-... .-- -n pursuant te Speeifieatien 6.6 prior te resurnptien ef plant epercltien. The written repart shall provide a deseriptior.

Cf investigatiens eendueted te determine the eauise ef the tube degradatien and eerreetive measures talzzr. te prevent reeurrenee.

In "TATT ýT?*T T T rTTyrm I

-7

,14 4 10e Amenenlment NC. 273

TABLE 4.4-1 XKlJ.4'R4TClff *'i '.1kfr5Mr nJJJ~

r~.1 .Jn.4. JJ 7%f. .. J.JMJýr',+/-..r7 JJ*l

!NSDECTED DURING PISElIVICEB INSPEECTIGN Preserviee inspeetien. We Yee-a N-. ..

mf Steam n.rater. per- Unit Three Three First inserviee inspeetien Aq _______________

Seeend & Subseefae-nt Inserviee inspeet-i-en Gne-44-) ______________

Table Netti-en"-

(1) The inserviee inspeetien may be limited to one steamn generater en a rotating seheduloe eneapassing 9 poreent of the tubes if the results of the first or previeu' inspeetions indieate that all steam generators are performing in a like manner. Nete that under seme

-. -umstanees, the eperating cenditiens in ene or mere steam generaters may be feund to h mere severe than these in ether steam gonerators. Under sideh eiretumtanees the saple sequoncc shall be m.dified to inspe.t the mest severe -ondition'.

(21~~~~~~~~~~~~~~~~~.

et.her The .6;;.fne-a- ne npee uie.-efrtisrie f-einsalb

ýt-

-sneeteA'~ mlit th1i~rd A-md -!hqtnc-rt inrtin hetidl felle; the inrstrucic Adeserinoi +-n (1) above-.

PRVAXMI VALLPT! T;TMTT I1 3/4 4 19d GP

STEAMN CENERATGR TUBE HISPECTION 1ST SAMPLE iNSPEC-TiO 2ND SAMPLE RTDPSGCT- iG 3RD SAMPIE; INSPEC-TiO Sampl-e Result Aetienr Reuirza .Result Aetien Reqaird Result Action leegired Ami-4n NeIu*e N/-A N-A NI-A

,e- S tubes G--21 Plug defeective Guz-______ Nete N/-A I NI-A per S.. rand inspect "dditienl G-4 Plug defective tee and C-1-! Nene 2S tubes in this S.C. inspect additional 4S tubes C- Plug dfcetiv' in thi SC. t-ibef G Perform actien.

fer C 3 result of first sample Perform action fer C 3 N-/-A

_________res~ult ef first sample ________

G-3,- Irspeet: aill tubes i~n All ether Nene NI/-A NI/-A this S.G., plug".9 a.

defective tubean d.

inspect 2S tubes in each ether S.C.

S.m. S.CG. Perform action fer C 2 N/-A N/-A Notifieti. te NRC are G.. 2. result Cf second samplC Pursuant but*" ne*-

Speeifieatien 6.6 ad~ditiefnal Additiernal inspeet ail tubesi in eaebh N-/A NI-A S.G. ia S.C. and plug defective C-4 tubes. Netifieatien te NRC pursuant te Speifitin*'

- (3 % '0 un-rom *; i-:; th :m nt:r et steara generatcrs J

inspneteel idurancR aR in nnrtinn.

-R T'OT97ý--jj- TT7NT T T" T"TIrm 3/4 4 1 3-/A-4-4GeAmenelment Ne. 273

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CON ITON FOR OPERATION

'2 A r t-nhp *jntpcrritv ~hF111 h~ m~iintained qQ t"'hp intparity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be pluaaed in accordance with the Steam Generator Program.

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

ACTT~Q1k

-- -- - ----------- - - - - 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 plugaed in accordance with the Steam Generator Proaram:

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

b. With Action a not being completed within the specified completion time or if SG tube integrity is not being maintained, 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 the Steam Generator Program.

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

BEAVER VALLEY - UNIT 1 3/4 4-8-- Amendment--N-Q--

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 I any one steam generator, and

d. 10 gpm identified LEAKAGE.

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

ACTION:

b~a. With any Reactor Coolant System operational LEAKAGE geater than

.n.any .f the abeve.not-withinlimits, .... udi-ng'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 er be in at least HOT STANDBY within the nex~t 6 heurs and in COLD) S!!UTDOWN% within the fellewing 30G heurs.

ab. With the recauired action and associated completion time of Action a not met, or with ent-pressure boundary LEAKAGE. or with primary to secondary leakaae 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-fo_30 hours.

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:(I)

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. 14&

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 )last once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steaey state 0-eaten. -I

c. Verifying primary to secondary LEAKAGE is less than or equal to 150 gallons per day thrpugh 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 MPODE 3 er 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a establishment of steady state operation.

(3) Not applicable to primary to secondary LEAKAGE.

BEAVER VALLEY - UNIT 1 3/4 4-14 Amendment No. 1-8 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 followina completion of an inspection performed in accordance with the Specification 6.19. Steam Generator (SG) Program. The report shall include!

a. The scope ofjinspections performed on each SG.

b. Active degradation 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 plugaed during the inspection outage for each active degradation mechanism.

f. Total number and percentaae of tubes oluaaed to date.

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

h. The effective pluaaina percentaae 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. 2-5, 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 Proaram shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Proaram shall include the following provisions:

a. Provisions for Condition Monitorina 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 1

,J 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 plugaging of tubes. Condition monitorina assessments shall be conducted during each outaae during which the SG tubes are inspected or plugged, to confirm that the performance criteria are beina met.

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

accident induced leakage, and operational LEAKAGE.

1. Structural integrity -erformance criterion: All in-service steam aenerator tubes shall retain structural intearity over the full ranae 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 Dower operation primary-to-secondary pressure differential and a safety factor of 1.A against burst applied to the desian basis accident primary-to-secondary pressure differentials. Apart from the above reguirements, 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 leakaae 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 leakaae rate for all SGs and leakaae rate for an individual SG.

Leakage is also not to exceed 1 apm per SG. except during a SG tube rupture.

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

.. Provisions for SG Tube Repair Criteria Tubes found by inservice inspection to contain flaws with a depth e -ual to or exceeding 40% of the nominal tube wall 1-hi ~

rý---- nl 11 hm

' r-1iirvTen9-A fý BEAVER VALLEY - UNIT 1e -= Amendment--N-o-

I, *b 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 obiective of detecting flaws of any tvpe (e.g., volumetric flaws, axial and circumferential cracks) that may be present alona 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 Dart of the tube. In addition to meeting the reguirements 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 intearity 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

_,k4 ,1 1, +- i-k~l ", e +- r% "kms = I r ml A n +-* r.7'h =

locations-

1. Inspect 100% of the tubes in each SG during the first refueling outaae 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 seguential period shall be considered to begin after the first inservice inspection of the SGs. During each period inspect 50%

of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outaae nearest the end of the period. No SG shall operate for more than 72 effective full power months or three intervals between refueling outjaag (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 interval between refueling outaaes (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or enaineering 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 Amendment-No-

Attachment B-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:

Page V

XIV XV 1-3 3/4 4-11 3/4 4-12 3/4 4-13 3/44-14 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 3/44-19 3/4 4-20 6-22 6-22a*

6-27 6-28*

6-29*

6-30*

6-31*

6-32*

New page

- Markup of page pending in LAR 173 (EPU)

,b 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 Integrity.......... 3/4 4-11 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. *4. I

- I I dated July 11, 2002. I

V.

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 REPORTIN G 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 --.

Steam Generator Tube Insneclinn Renort. 6-2-2 6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM ..................... 6-22a I BEAVER VALLEY - UNIT 2 XIV Amendment No. 14-9

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 (Liauid, 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) Proaram .......................... 6-27 BEAVER VALLEY - UNIT 2 XV Amendment No. 1-34

I .

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

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 s-team g.n.rater tub"e nrimary to secondary LEAKAGE) through aj nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

BEAVER VALLEY - UNIT 2 1-3 Amendment No. 9-

I IS I&

REACT-T COOLANTq- SY,&T-NS 3ý.. A STEA cERA.7RS LIMiITING COENDITION FRo OPERATIO-N 3.4.5 Eaeh setxam gcneratzr shall be OPERABLE.

APPLICABILIT-Y. MGDES 1, 2, 3 and 4.

With ene er znrce siteam generators~ inoperable, resitore the ineperable generator (s t...... L..... status prl.r n..-

..... i,, T apove ' ......

e'Tm TT**' T T *% FI* *ITTT ,%R*%TT n 4.4.5.1:

h*_'* . s...... Steam J~

...... ~J~J.L.J Ccnerater L..4 A.6. Sample Selectien and lnnep-ctien Each stca!

generater shall be determined OPERABILE Adurilng shutdewn by seleeting and inspecting at least the minimum number ef steam generatzre Zspeeified in Table 4.4 1.

4.4.5.2 Steamn Concrater Thdbc Sample Scleetien and inspeetien The steamn gcncratzr tube mainimu Sam~ple Size, inspeetien result classification, and the eerrcspending aetien required shall be as specified in Table 4.4 2. -The inserviee Inspeetien ef steam generatzr tubes shall be perfermed at the frencc ~peeified in Specifieatien 4.4.5.3 and the inspeeted tubc3 cvrified ehl acccptable per the aeeeptancc criteria ef Speeificatien 4.4.5.4. Steam generater tube-a shall be examined in aer-danee with Artiele 8 Cf Seetien V ("Edd Current Emaminatien ef Tubular Preduets") and Appcndim 1V te Sectien X! ("Eddy Currcnt Emaminatien Cf Nenfcr-romagnctie Steam Ccncrater Hecat Exchanger Tubing") of the applicablc year and addenda Cf the ASME Beiler and Pressure Vessel Cedc required by lOCERSO, Scctien 50.5sa(g) . W~hen applying the cecccptiens Cf 4.4.5.2.a tr- g 4.4.5.2.c, prcvietus defect- cr . ..mpC"-fetiens in the area rep-io-* by sleeving are net eansidered an area rcirn reinspeetien. The tubes scleetcd fer cach inservicc inspeetien shall include at least 3 pcreent Cf the tetal: number Cf tubes in all steam generaters;th tubes seleeted far- these inspeetiens shall be seleeted en a randein basis emcopt:

a. W~here experiencc in similar plants with similar water ehemistry indicates eritical areas te be inspeeted, then at least 50 perccnt Cf the tubes inspeeted shall be frem these critieal areas.

b. The first sample Cf tubes seleeted fer caehinrvc inspcc[lon 3sutbsequent t:. the pr- :scrvce ispeeti~fl or-cach stceamf generater shall ineludc.

BEAVER VALLEY UNIT- 2 3'4 4 ii 3L44AcdctN.5 Amendment Ne. 52

I.

711"7ýr-m^n n"ItIm"Am SURVEILLANCE REQUIREMENTS (Gentinued)

i. All rnanplugged tubes that previeusly had datactabla wall penetratienn greater than 20 peraarnt, an

2. Tubes In these areas where exper-lenee has indieated patantial prebleamn, and

3. At least 3 pereent af the tetal numfbar ef sleeved tubes in all: :three nteaam ganeratern. A sample niz-e less thant 3 pereent is aeeaptable prevideed all: the sleeved tuben in the steaam generater~s) ex~amined during the refuaeling eutage are innpaated. These inspaatiens will inaluda beth the tube andl the sleeve,

4. A tube innpeetien pursuarnt ta Spaaifieatian 4.45.4.a.8. If an~y nalactad tube deen net parmait tha pasnage ef the addy Current probe fer- a tube er sleeve innpaatien, this shall be raaardad and an adjaeant tuba shall be naleatad and subijaatad ta a tuba inspaatien.

S. indieatians left in sarviaa as a result af applieatian af the tuba nuppart plate valtaga based rapair'-

ritaria (4.45.4.a.19) shall be inspaatad by ba~bbir.

e+/-il praba during all future raefualing autagan.

e. The tubes nalactad an the saaand and third namplan (i-f-required by Table 4.4-2) during eaah innarviaa inspactian may be subjaatad ta a partial tuba inspactian pravided:

1. The tubes salaatad far these samples inaluda the tubes framf these areas af the tuba sheet array where tuba with imparfaatiann were praviausly fauind, and

2. TFha innpaatianns inaluda these partians af th tua where imparfaat: HFIR WC-Pr-P prev1QueJly fetuncl.

di. implamaintatian af the nteaam ganaratar, tuba- ta- tuba nuppart-iinnpaatian far hat lag and eald lag tuba nuppart plate irntarnaatiann daewn ta the lawant aald lag tuba nuppart plate with knawr. autnida diameter stress aarraniar. araaking (ODZCC) indiaatiann. The datarmainatian af the lawest aald lag tuba nuppart plate intarnaatiann having 9EDSCC indiaeatiann shall be banadl an the parfarmanaa af at leasnt a 20 paraaent r-andamf naffpling af tubes innpaatad ever their-full length.

P=UI79P UnT T RV TWT?'JT p 3/4 14-44 12 - - ý IKIT- -1 n -I

SURVEITLL7NGE RlEQUIRlEMENTS (Continued)

The Eesults of cach sample irnspcction shall be classified inte ene of the fellewing thrcc categorcs G-4 Less than 8 perccnt of the total-tubes inspectcd are Elegraded tubes and nono of the inspeeted tubes are defcctive.

G--Z GOnc or moerc tubes, but net moer-than 1 pcrccrnt of the total tubes inspcctcdl a re defectivc, orI between :5 perccnt and 10 perecnt of the total tubes inspccted arc-elegraded tubes.

G-4 Mere than 10 percont of the totalr tubes inspccted ar-e cgradcd tubes or moro~ than 1 pcrccnt of the irnspcctcd tubes arc dcefcctiveo.

Nete~s in all inspcctions, prcviously elcgradc tubes or sleeves must exhibit significant Cgrcatcr than 10 perccent) fur-thcr wall pcnctratieris to be incluided in the aboyc pcrccntagc calculations.

4.4.5.3 linspctiern Frceauancics The above recriredinric inspectiens of steam gorncrater tubes shall be performeel at thes following Erefrcencics:

a. The fir-st inservicc ins cctien shall be pcrformcd after 6 Effcctivc Full Pewcr Months but within 24 calendar months of initial criticality. Subsequent inscrviec inspcctiens shall be pcrfermcd at intervals of not less than 1:2 nor mere than 24 ealcndar months after the prcvieus inspcctien.

!r- two eensccuitivc inspcctiens rtoilowing scrvicc tinder All Volatilc Treatment (AVT) eonditions, not including the prcscrviec inspcctien, result in all inspcctien results falling into the G 1 catcgery or if two conseetutive inspcctiens dcmenstrate that prcviously obscrvcd dcgradatien has not eontinucd and no additional dcgradation has occuirred, the inspection interval may be extended to -a maII-t- o f onco per 40 moinths.

BEAVER VALLEY MUT 2 "9 /4L4 4 4:3

I, i tt~ft.= i. U t k.AJULLJtf+/--b+/-bt +/-!3M.LI SURvE iLL2iC-E REQUIREIIENTS (Centinued)

b. If the inserviac in.zpti.n ef a steam gnratr ,.nduet..

In a.5cc datnc with Table 4.4 2 requires a third sample infp istirnwh,,.results fall in Cat-g.ry G 3, the inspeetien freefuaney shall be inereaacd te at least enee per 20 menths. The inereaas in inspeetizn fraquency shall apply until a subsequent inspeetien defmtnstrates that a third sam.ple inp*."ti.n is net required.

e. Additienal, unschadulad inserviee inspeetiens shall be perfer-med en eaah steoam genernater in accar-danca with the first sample insp.cti.n s . in Table 4.4 2 during the

.ifi.d shuitden subsequent te any ef the fellewing cenditiens.

1. Primary te seeendary tube leakcs (net ineluiding leaks e-riginating frem tube te tube sheet welds) in emcaaa ef the limits ef Speeificatien 3.4.6.2,

2. A seciamic Cccuirrenee greater than the Operating Basis Barthquake,~

3. A lessI Cf eeelant accidant rurigaetuatien ef thae anginaara--d safeguards, er

4. A main steamline er feadwater- line breaak.

4.4.5.4 Accaptanca C-ritearia'-

a. As used in this Specifieatien.

1. impaerfectien means an emeeptien te the dime.n.i-nS, finish or contour Cf a tube or sleeve from tha required by fabrieation drawings Cr specifieatien-s.

Eddy current testing indicatiens below 20 percarnt af the nominal: tube Wall thickness, if datactabla, may be eensielered as imperfeetiens.

2. Dee~radatien means a sarvica inducad crackeing, wastage, wear Cr general c .rein Cccurring efl either insid Cr Cutside Cf a tube Cr sleeve.

3. Daecradaed Tutba malans a tube er sleeve eentaining emprfeetiens greater than er equal tC 20 parcant Cf taneminal: wall: thiclenass causad by dagradatien.

P, "A 71m l I .A T.T. . .v T .MTTr'P 3/4-4-14 .............

m o. 1:01:

'I IT .Proposed changes to draftpage

ý t. . .TlU. t . UL..1 L= t1. LbIb l f..

&i from Unit 2 LAR 173 (EPU)

'r T T 'A ?ý " "T, "Ir""Mlmamn tý 4- , A muflv Vý= S2LJ4 1A "= I-= d 7~S

4. Pareent Deeradatien Faeans the pareantaga ef the tube er sleeve wall thiekness affactad er ramevad by dagraelatien.

S. Dafeet means an imperfactien ef sueh severity that it

1. £ ~ f/~U I. '... - ..I- U 4-ILt.

cantaining a daefeet is defeetiva. Any tube whieh dee-s net permit the passage of the addly eurrent inspact-ian preba shall be deeamad a dafactiva tube.

V

at Or biyind whiah the tube shall: be remva_4 fre sarviea by pluggi~ng ar raepairad by sleeving in th affactad area beaesue it may bacama unsarvieaable priaer te the next inspeetien. The plugging er repair limit imparfactien depths are spacifiad in pareantaga af neminal wall thielenaca as fellews:

a) Original tuba wall 4096

[Phis dafinitian deaz net apply ta tuba stuppart plate interseetiens fer whieh the valtaga based repair eriteria are being applied. Rfeafr te 4.4.5.4.a.1:0 for the repanir limiLt applieabla te these intermetin. .. .

b) ADD Gembustien Engineering TIC welded 2796 sleeve wall e) Westinghetuse laser walelad sleeve wall 25%

7 TUnservieeabl- d.s "riba. the .

..nditian f a tuba if it leaaes or eentains a defeet large eneugh te affect its atrutural integrity in the avant af an Oparating Basis Earthquake, a less ef coolant accidant, ar a staamline ar faadwatar line breaak as spacifiad i 4.4.5.3.e, abava.

8. Tube Innapactian means an inspactien af the stae ganaratar tuba fram the paint af entry (hat lag side) eamplately araund the U band ta the tap suppert ta the eeld legr P=Irp - TTAT 7 TbXTE33 P 3/4 4 l4a,

4.LAmad*inant N..

1, 1.

,UR39TEIL3ANCE REQUIREMENTS (,antinu",d)

I from Unit Proposed ch2anges LAR to173 (EPU) draft page

9. Tube Reva-Lir refers te sleaving whieh is used te mnaintain a tube in sarvica er return a tube te serviea. This includas the ramaeval ef plugs that were installed as a carractive er preventive maeasure. The faellawing sleeve designs have been faund aeeeptable.

a) ABB Cainbustien Engineering TIC welded sleeves-,

CEN 629 P, Revisien 02 and GEN 629 P Addendum 1.

b) Westinghetuse laser welded sleeves, WGAP 1348-3-,

flavisien 2.

10 . TubrSutpart Plate Plugging Limit is used f-r the dispeasitien af anallay 60 stea.a gen.rat.r tube f.r cantinuaed servica that is emper-i-cing predaminantl-y axially eriented eut-sida Adiameter stress carrasi.an eracking canfinad within the thickness ef the tube suppart plates. At tuba suppert plate intarsactians, the plugging (repair-) limit is based en maintaining steam garnaratar tuba sarvieaability as dascriba a) Steam, generatar tubas, wh.s ,dagradatian is attributed to autsida diameter stress carra---

cr-a~cking within the baunds af the tuba suppart plate with bebbin valtagas less than ar equal ta 2.0 valts %ill: be allewad La remain in sarvica.

b) Steam ganaratar tubas, whasa degradatia~n.i attributed ta autsida diameter stress carrasia craecking within the baunds af the tuba suppart plate with a babbin valtaga greater- than 2.0 volts will be repaired or plugged, except as

nrj~~p j:M 44511OjA;,;u rýT' 7kT ~r, !T7VTTI ¶7"A TThTT-M 7 L4)4 3/4 4 9:4b- Amendment Nli.

gL SURVEILLANCGE REQUIREMENTS (Cont inued) e) Steam generater tubes, with indieations ef petential degradation attributed to eutside diameter stress . .rres..n .ra.king within the bouends of the tube .upp.rt plate with a b.bbin*.

v.ltage greater than 2.0 velts but le* o than er eequal te the upper v.ltage repair limit-+---mya rominin servico if a rotating pancake coil or

..cccptable alternative irnspeetien docs not detect degradataion. Steam generater tubes, with indications of eutside diameter stress corrosion eraeking degradatien with a bebbin vtaogreater than the upper veltage repair lmt ilb plugged or repaired-.

d) If .

an unseh-dul"d mid .y.le inspeetion is 4 rfer5 d, the following mid -yelerepair linits apply instead Cf the limits identified in 4.4.5.4.a.!O.a, 4.4.5.4.a.!O.b, and 4.4.5.4.a.!O.e.

The mid cyclo repair limits are determineod from tho-following oquations.

v *T 1.0+ NDE +Gr )

CL

_,, , ,, (CL-At]

VMLRL - -M(L"VuL-~ "LRL/ )(LA)

. T ut %r. CL (1) The upper voltage repair limait is calculated according to the methodology in Conerie Letter 95 05 as supplemented.

BEAVER VALLEY UNTR42 3,/-4 4 4--4e Amendment No. 10-1

RFACT-GR CODL3AIT SYSTEM SURVS!1LLA;C-E RBQUIREmENTZP (Coentinued)

"T'rl* T* n-T "I" "* %7t'1*"* T*)T*f'%T'r"I" T* *'l?*,Tt"'ATrn C1 .-1\

upper veltag repair S

limit

_________________ H - lo

!wer voltage repair limit I t "Ift * .*

VmIURL-b --mid eycic upper veltage repair lim.it based en time into .y..e VIILL - id eycic !ewer veltage repair limit based n*

VmuRb and time into cycle At - length of time sinc last s*h*duled inspection during whi"ehl GL -3 cyel length (the time between twe seheduled steamn generater insp*otioenz)

____________________ structural limfit voltage Cr - average gr.wth rate per cycle lng-th NDE - 95 percont cmltv probability allewanee for nondestructivo (i.e., a value of 20 1--- ý as Deen approved by NRG) 4-2 implementation 5yle of these mid repair limits sh5uld.

follow the same approach as in TS 4.4.S.4.a.l9.a=,

4.4.5.4.a.10.b, and 41.1.5.4.a.!O.e.

4. 1 . £Aýý wyAm .a L.CA' V= ml frJJ.. '.. 1,A MAVI 4 -N 1.J -T n' .

supplemented.

7 Prý'A3=Tý XZAT..T.RW MITEP p 3/4 4 l4d TF'.. ~hT~

S~7T 2 ~L444dAmeondmeont No. 101=-

WL RFlACT-GR COO3ANT- SYSTEM ib uUXL+/-L+/-2JKS EQUIREMENTS . t Pone lttl.a.)

b The steam gen.rat. r shall be determined -PERABLEaft..

eempleting the eerrespendirng actioens (plug or repair all tubes aercccding the pluggin~g or repair limit) required by Table 4. 4-,2.

4.4.5.5 art

a. Within 15 days fellewing the completior. of eaahinric insp1ctirn of steam generatr tubes, the numbr of tubes plugged or repaired in cach steam generater shall be submitted in a Speeial flapert in accor-danca with 10 C-F

b. The eemplete results of the steam generater tube and sleeve inservica inspeetion shall be submitted in a Special Reaport in accordanca with 10 CFR 50.4 within 12 months following the eompletion of the inspection. This Special Report shall include.

1. Number and exctent of tubes andl sleeves inspeeted.

2. Location and peernt of wall: thickn.ss penetration for eaah indipatien ef an impfrf..tion.

B. idantifieation of tubes plugged or repaired.

e. Reasults of steam ganarator tuba inspections which fall into Category G 3 shall be reported to the oiso pursulant to Spe"ifi"ation 6.6 prior to rosumption of plant

-prati.n. The written rp*ort shall provida a daesription of invastigations eonductad to determine the eausa of the tube dagradation and corractiva measures talean to prevent racurranea.

d. For implamalntation of the voltaga based repair critaria to tuba support plate intarsactions, notify the Ge....zs----

prior to raeturning the steam ganarators to sarvica (14ODE 4) should any of the following conditions arisa

1. ,.,,i.

If estimated

,,JJL%*.%,,..I,.

!* 4A,.-P leakage 4-. - J%,,.*. -

based 4-4*%,,4 %.

on the4-,J., projactad vmýI[*4.

4. J. -. [,ý . - %m%.

4- -

1 %.4 andl ofJ.

. ^4 ... *1 '1 4 . - - __ -4 * * *.. - .. - .. . ., .i-,t,., "1 and of c1*cl) voltaga distribution em-eeds the leak imffit (deter-mined from the licansing basis doesa calculation for the postulatad main steaam-line breaak) f~p.r 4ihq Ap.". oparating cyela

]BEAVER VALLEY U-aL4.e IWTTT'l 2 3ý4-4-l-4e

1. V" REACTOR ' G-LA' T Y*rTE'R ZIJRVEILLANjCE REQUIREMENTS (Ccntinued.)

2. if eireumrferential eraeck like indieaticns are deteeted-at the tube stippert plate interseetiens.

3. if indieatiens are identified that extend beyend the eenfines of the tube stippert plate.

4. if indieatiens are identified at the tube stippert plate elevatiens that are attributable te primary water stress eerreslen cracking.

S. if the ealetulated eenditienal burst probability based en the prejected end ef cycic (er if net practic-al, tusing the actual measured end ef cycic) veltage distributien emeed 1 X 1:0 , netify the Gelt-i-----

and previde an assessment ef the safety significancc T-T T4 rqT1 -

l 1Pr.,AMP, ZTATT.Tz.P TD.rTT P.1 3ý4 4 14-f- Amendment Ne. 1:01

1.

T'ABL3E 4-.4--4 14114NWMq! NTJBER OF- STEAMŽ GENERATOGRS TO BE INSPECTED DURING INGERVICE INSPECTIIOGN Prrcsrviee inspecti~n. -NeYe Ne. ef Steam Ccneraters per Unit -Thr-ee -Thr ee First inserviee inspeeti~n. -M4Tw Seeend &. Subsequent inserviee inspeetiens _-Gný-ele Table Net-at-i-n

1. The inserviee inspeetien may be limited te ene steam generater ena retating sehedule eneempassing 9% ef the tubes if the results ef the first er previeus npcin indieate that all steam generaters are perferming in a like manner. Nete tha-t under seme cireumstanees, the eper-ating eenditiens in ene -r-mere steam generaters may be found te be inere severe than these in ether steam generaters. Under stieh eiveuintancczi the sample seu~cC ene hall bez medified te inspeet the mest severe zndiitiens.

2. The ether steam generator not inspected during the first Inserviee inspeetien shall be inspeeted. The third and sub eguent inspeetiens shetuld fellew the instruetien dclseribed-T't* IN~

T)*%*TT*f% T Tr 11- *

vl- 91" Tv T 3,/4 4 1:5

TABLE 4.4 9T-EAM CENERAT-R TUBE !NSPEC-TTION 1sT- SAMPLE R;SPECTPiN 2ND SAMP1LE INSPEC-TIGN 3RD SAMPLE IN;SPECTIO Samp-le--e* Result Action Rfe -ir-d Res4 Action Reoquired Res-l Actior.n fqu"ir.d A mainimuN G1 N4-A N/-A NAk Of S tubea G-4 Plug .r repair G-4 Nne NW-A N-/-A per S.__ _____--__

tubes and. inspect G Plug . r repair d" fetive G-1' Neon additional 2S tubo" in tubes and in"pt additional G-4 Plug r p ti .. 4S tubes in this S.C defoctive tubea G Perform aetion for C 3 result of first safp-le Perform action far C 3 N*-A Ný-A

______________________result of first seoT-lo _______

G insJpcct all tubes in All-ether Nene N-/A N-1/A this S.C., plug rS..s are repair defeetivo tubc G--l and inspeet 2S tubef________________

in each ether S.C. Seine S.C.a Perform action for C 2 N-/A N-/-A C 2 but no result of seeend sample Notification to RC additional-pursutant toG.s are Specification 6.6. __._

Additional- inspect all tubes irn each. N-/-A N-WA S.Gis S.C. and plug or repai Gdefective tubes.

Notification to NRC pur....nt.

____ ____ ____ ___ _ __ ____ ___ to Specification 6.6-. _ _ _ _ _ _ _ _ _

? - , I~ -- - V- 3a ir nm r11,mr=r-m n r -fr1T : -

-Fr.-~- - -u ~- crrr I . -

PRAUFR XAT.¶ TUTUD~ !D7444 3ý4 4 16 mndnn Ne.

Amenelment o 0 1:01:

REACTOR COOLANT SYSTEM 3/4.4.5 STEAN ONERATOR (SO) TUBE INTEGRITY LIMIITING C NDITION FOR OPERATION I A

r S( tube *intparitv shall be maintained SG tube intearity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plucaed or repaired in accordance with the Steam Generator APPLICABILITY: MODES 1, 2. 3. and 4.

ACTI-QIN

.-.-.-. .GENERAL NOTE Separate action statement entry is allowed for each SG tube.

a. With one or more SG tubes satisfying the tube repair Steam Generator Program:

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

2. Plua 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

b. With Action a not being completed within the specified completion time or if SG tube integrity is not being maintained, 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 REOUIREMENTS 4.4.5.1 Verify S9 tube integrity in accordance with the Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies the tube repair criteria is Dluaaed or repaired in accordance with the Steam Generator Proaram prior to entering MODE 4 following a SG tube BEAVER VALLEY - UNIT 2 3144-11-- Amendment-No-

%, 4(

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 .n. ef the...vclimits, az........

reasons other than pressure boundary LEAKAGE or primarv to secondary LEAKAGE, reduce the LEAKAGE rate-eto within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months er be in at least 1OT STAiDBY within th-next 6 heurs and in COLD) ZIUTDGM4 within the fellewing 30 heurs.

ab. With the reauired action and associated completion time of Action a not met, or with any-pressure boundary LEAKAGEo with primary to secondary leakagle 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. 14)71-

'S 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 stea.dy state 193, At Ae , (2) (3)_

c. Verifying -rimary to secondary LEAKAGE is less than or eaual to 150 aallons per day through any one steam aenerator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .(Z)

(2) Not required to be performed in MODE 3 er 4 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. -64 1

a. a 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

1. 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.

b. Active degradation mechanisms found.

c. Nondestructive examination techniaues 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.

q. 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.

2. A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.19. Steam Generator Program. when voltage-based alternate repair criteria have been applied. The report shall include information described in Section 6.b of Attachment 1 to Generic Letter 95-05. "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking ."

3. For implementation of the volta-ge-based repair criteria to tubesnupport plate intersections, notify the Commission prior to returning the steam generators to service (MODE 4) should any of the following conditions arise:

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

BEAVER VALLEY - UNIT 2 6-22 Amendment No. 4 il ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (continued)

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

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

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 shall 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 Amendment-NQ-

'a 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 followin *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 Inteority SG tube integrity shall be maintained by meetino 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 integritv over the full range of normal operating conditions (including startuo. operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and. except for flaws addressed through application of the alternate repair criteria discussed in Specification 6.19.c.4, a safety factor BEAVER VALLEY - UNIT 2 6 -27 Amendment No. 4-24 1

I. IS ADMINISTRATIVE CONTRQOLS STEAM GENERATOR PROGRAM (Continued) of 1.4 against burst apolied 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 sirni ficantl affect burst or collapse shall be determined and assessed in combination with the loads due to oressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

When alternate repair criteria discussed in Specification 6.19.c.4 are applied to axially oriented outside diameter stress corrosion cracking indications at tube support olate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than IxlO-L

2. Accident induced leakaqe__erformance criterion: The primary to secondary accident induced leakage raLefD-r 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.

Except during a steam generator tube rupture. leakage from all sources excluding the leakage attributed to the degradation described in TS Section 6.19.c.4 is also not to exceed 1 gam per SG.

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

c. Provisions for SG Tube Repair Criteria

1. Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth eaual to or exceeding 40% of the nominal tube wall thickness shall be olugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in Specification 6.19.c.4.

I.

2. Tubes with sleeves found by inservice inspection to contain flaws that are not in the sleeve to tube ioint. with a depth eaual to or exceeding the following percentages of the nominal sleeve wall thickness. shall be oluaaed:

ABB Combustion Engineering TIG welded sleeves 27%

W2wtqhouse laser welded sleeves 25%

3. Tubes with a flaw in a sleeve to tube joint shall be

4. The following alternate tube repair criteria mayv be applied as an alternative to the 40% depth based criteria of Technical Specification 6.19.c.I:

Tube Support Plate Voltage-Based Repair Criteria Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube f-or BEAVER VALLEY - UNIT 2 5=2a Amendment No.

t& p 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 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 olugged.

except as noted in 6.19.c.4.c below.

c) Steam aenerator tubes, with indications of potential 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 indications of potential 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 accordino to the methodology in Generic Letter 95-05 as supplemented) will be plugged or repaired.

BEAVER VALLEY - UNIT 2 6-29 Amendmen

S 1S ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) e) If an unscheduled mid-cycle inspection is performed.

the following mid-cycle repair limits apply instead of the limits specified in 6.19.c.4.a. 6.19.c.4.b.

6.19.c.4.c and 6.19.c.4.d.

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

V*T VML MURL =

CL At) 1.0+NDE+Gr( ( 7L VMLRL = VMURL -(VURL - VL) CL )A whex2eL YURI, upper voltage repair limit YLR, = lower voltage repair limit mid-cycle upper voltage repair limit based on time into cycle mid-cycle lower voltage repair limit based on VMUR' and time into cycle length of t-i-me since last scheduled inspection during which VURL and2VLRL were implemented CL = cycle length (the time between two scheduled steam generator inspections) structural limit voltage average growth rate per cycle length NqDE = 95-percent cumulative probability 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.

Implementation of these mid-cycle repair limits should follow the same approach as in Specifications 6.19.c.4.a through 6.19.c.4.d.

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 Amendment--No-

I ADMINISTRATIVE CONTROILS STEAM GENERATOR PROGRAM (Continued) outlet. and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not Dart of the tube. In tubes repaired by sleeving. the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection. In addition to meeting the requirements of d.1. d.2. d.3. and d.4 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be oerformed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment. to determine which

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 oeriods 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 interval between 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 interval between refueling outages (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.

BEAVER VALLEY - UNIT 2 56-31 Amendment--NQ-

40 ,'

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued)

4. Indications left in service as a result of application of the tube support plate voltage-based repair criteria (6.19.c.4) shall be inspected by bobbin coil probe during all future refueling outaaes.

Implementation of the steam aenerator tube-to-tube support plate repair criteria requires a 100-percent bobbin coil inspection for hot-leg and cold-leg tube suoport 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-lea tube support plate intersections having ODSCC indications shall be based on the oerformance of at least a 20-nercent random samplina of tubes inspected over their full

e. Provisions for monitorina 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 intearitv of SG tubes without removing the tube from service. For the purposes of these Specifications. tube oluaaina is not a reoair. All acceptable tube repair methods are listed be~low

1. ABB Combustion Enaineerina TIG welded sleeves.

CEN-629-P. Revision 02 and CEN-629-P Addenduml-

2. Westinghouse laser welded sleeves, WCAP-l34830 Revision-2.

BEAVER VALLEY - UNIT 2 Amendment No.

I . 11 Attachment C-1 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:

Provided for readability only

Providedfor Information Only.

TECHNICAL SPECIFICATION BASES INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS ...................... B 3/4 4-1 3/4.4.3 SAFETY VALVES .............................. B 3/4 4-Ig 3/4.4.4 PRESSURIZER ................................ B 3/4 4-2 3/4.4.5 STEAM GENERATORS (SGI Tube Integrity........ B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ............. B 3/4 4-3 3/4.4.6.1 Leakage Detection Instrumentation .......... B 3/4 4-3 3/4.4.6.2 Operational Leakage ........................ B 3/4 4-3d 3/4.4.6.3 Pressure Isolation Valve Leakage ........... B 3/4 4-3j 3/4.4.8 SPECIFIC ACTIVITY ......................... B 3/4 4-4 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ............... B 3/4 4-5 3/4.4.11 RELIEF VALVES ............................. B 3/4 4-11 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 BORON INJECTION SYSTEM .................... B 3/4 5-2 3/4.5.5 SEAL INJECTION FLOW ....................... B 3/4 5-3 BEAVER VALLEY - UNIT 1 B-II Change No. 1-414= 1

REACTOR COOLANT SYSTEM Provided for Information 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.

3,4..A STEA GEy-NB~cAT4,r One GPERABLE steam gonerater in a non isolatod reacter coolant loop pr'des suffieient heat removal eapability to remeve decay hoa ftra reactor shutdewn. The requirement fer twe OPERALE stea generaters, combinod with ether- requirements of the Limit-ing Conditions for Operatior. ensures adequate deeay heat refmoval eapabilities for RCZ temperatures greater- than 35011F if one stea generater beeemes inporable due to single failure eensideratiens.

Below 3501F, deeay heat is removed by the PUIR system.

The Survoillanco Reefuiromoints for inspectien of the steam generater tubes ensure that the structural integrity of this portien of the flCS will be maintained. The program for inservico inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. inservico inspection of steam generator tubing is essential in order to maintain surveillanca of the conditions of the tubes in the event that there is evidonco of mochanical damage o progrossive dogradatior. due to design, manufaeturing errors, o insarvico conditions that lead to corrosion. Ir3cviee inspoction o-f steam generator tubing also provides a means of characterizingth nature and cause of any tube degradation so that eorreciv moasures ean be taken.

Tho plant is expectod to be operated in a manner such thatth socondary coolant will: be maintained within those parameoter limits found to result in negligible corrosion of the steam genorater tubes.

if the sacondary cool:ant chemistry is not maintained within these parameter limits, localized corrosio n may likely resualt in stress eerresien cracking. The extent of crackeing during plant BEAVER VALLEY - UNIT 1 B 3/4 4-2 Change No. 1-44-1-029 I

REACTOR COOLANT SYSTEM Providedfor Information Only.

BASES BASES 3/44. A TE4 GENERAT-GR (Gentin eperatizrn weuld be limited -by the limitatien ef steam generator tub leai age between the Primary Geelant Syostem and the ... .n.ary .la Systema (primary te secendary LEAKAGE w 150 gallens per day per steaa generater). Maintaining a primary te saccrndary L3EAKAGE less than this limit helps te ensure adequate mnargin te withstandl the izads imp"s.d during ne""al.peration and by postulated accidarts.

eperating plants have deamcnstrated that primary te secendary LEAKAGE ef 150 gallens per day per steeam generater ean readily be detected.

Leakage iin axcss ef this limit will require plant shutdsw and an unscheduled inspeetien, during which the leakeing tubes will: be lecatad and plugged.

Wastage type dafacts are unlikAcly with preper ehemistry ef seeendary:

coolant, su"h as pr.vid.d by All V"latila Traatm'nt (AýT).. However even if a defeet ef similar type sheuld develep in sarvica, it will be found dur*ng s.h.dulad insarvica st-e ge-nrater tube examinatiens. Plugging will be required Cf all tubas with imparfactiens axcaaeding the plugging limit. Steam ganerater tuba inspcctiens Cf eperating plants have d .m.nstrat.d the capability t-reliably datact a wastage type dafact that has penetrated 20 parcant ef the Criginal tuba wall thicknass.

W~henever the results Cf any steam ganarater tubinginrvc inspection fall into Category G 3, th*s* results will be r.p.rtad to the Gemi'sslen pursuant te Spacifieatien 6.6 prier te resumeption c plant aparatien. Such eases will be censidared by the GemmissieA Cn.

a ease by ease basis and may rasult i~n a raicatfer analysis, laberatery exetminatiens, tests, additienal eddy currant inspactien, and ravisien Cf the Tacehnical Spacificatiens, if nacassary.

3/4.4.5 Steam Generator (SG) Tube Integrity BACKGROUND Steam cfenerator tubes are small diameter, thin walled tubes _tha-t carry primary coolant throuah the primary to secondary_ heat exchangers. The SC tubes have a number of important safety functions. Steam aenerator tubes are an integral Dart 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 primarv coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unigue 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 SC 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).

ii S I .Provided.for Information Only.

SG tube integrity means that the tubes are capable of performin' their intended RCPB safety function consistent with the licensing basis, includina applicable regulatory recguirements.

Steam aenerator tubina is subject to a variety of degradation mechanisms. Depending upon materials and design, steam aenerator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion crackina. alona with other mechanically induced phenomena such as dentina and wear. These dearadation mechanisms can impair tube integrity if they are not manaaed effectively. The SG oerformance criteria are used to manaae SG tube dearadation.

Specification 6.19. "Steam Generator (SG) Program." reguires that a program be established and implemented to ensure that SG tube intearity is maintained. Pursuant to Specification 6.19. tube intearity 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 intearitv at normal and accident conditions.

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

APPLICABLE SAFETY ANALYSES The steam aenerator 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." olus the leakaae rate associated with a double-ended rupture of a sinale 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 discharaed throuah the main condenser.

The analysis for design basis accidents and transients other than a SGTR assumes 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 leakaae limit of 150 gpd per Sr. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EOUIVALENT 1-131 is assumed to be eaual to the LCO 3.4.8, "RCS Specific Activity," limits. Pre-accident and concurrent iodine spikes are assumed in accordance with applicable reaulatory guidance. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaaed fuel. The dose conseeauences of these events are within the limits of 10 CFR 50.67 as supplemented by Reaulatory Guide 1.183.

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

LCO I Provided for Information Only.

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

During an SG inspection, any inspected tube that satisfies the Steam Generator Proaram 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 lenath 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 Proaram also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SP 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 aross 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 reguires 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

I I Providedfor Information Only. I 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

Dearaded Steam Generator Tubes,

" August 1976.

The accident induced leakaae 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 exceed 150 apd per SG. The accident induced leakaae 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 aallons per day. This limit is based on the assumption that a single crack leaking this amount would not propaaate 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 challenaed when the pressure differential across the tubes is larae. Larae 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 reguired actions may allow for continued operation, and subsecuently affected SG tubes are governed by subseauent condition entry and application of associated reguired 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 pluaaed in accordance with the Steam Generator Proaram as reauired by SR 4.4.5.1. A-in evaluation of SG tube integrity of the affected tube(s) must be made. Steam aenerator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Proaram. The SG repair criteria define limits on SG tube dearadation that allow for flaw arowth between inspections while still providing assurance that the SG pmrfcnTmnnrp c-ritprin will rcntimnle to hp mpt. Tn order to

Providedfor Information Only.

determine if a SG tube that should have been pluaaed has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outaae 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 arowth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Acion plies.

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 refuelina outaae 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 followina 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months s-The allowed completion times are reasonable, based on operating experience, to reach the desired plant conditions from full -ower conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 44.5.-l Durina shutdown periods the SGs are inspected as reguired 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 Proaram 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 monitorina 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 Proaram in coniunction with the degradation assessment determines the scope of the inspection and the methods

I Provided for 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 dearadation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Proaram defines the Freauency of SR 4.4.5.1. The Freauencv is determined by the operational assessment and other limits in EPRI. "Pressurized Water Reactor Steam Generator Examination Guidelines." The Steam Generator Proaram uses information on existing degradations and growth rates to determine an inspection Freauency 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 reauirements concernina 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 arowth. 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 Freauency 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 nluaaed prior to subjectina the SG tubes to significant primary to secondary pressure differential.

BEAVER VALLEY - UNIT 1 B 3/4 4-2a Change No. I-Om'4M

REACTOR COOLANT SYSTEM I Providedfor Readability 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

REACTOR COOLANT SYSTEM RLProvidedfor 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 ProvidedforReadability Only.

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

I' REACTOR COOLANT SYSTEM I Providedfor Information Only.

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 throuqh Any One SG I eperating ac-periarnaa at PWR plants has shewn that sudden incrcascs in l*a*r rate are ef t-n pr..urs.rs t, larger tube.

f ailuras. Maintainin~g an eperatirng LEPOKPCE limit ef 150 gpd per steam ganaratar will minimfiza the patantial far a large LEAKAGCE event at pow1r. TPhis perating LEdAy eG limit is mere restrietive than theA peratinrg LEAcriE lim in standardized t9"h-i0al speaifiPatimns. This prevides additional margin tN a9-atemdats a tubeor flaw hio h might grew at a greater than expectgd rate Gr tianbe tdly extednd eutside the thieknass ef the tuba suppart plate= This raduad LEAKACE limit, Th in nntit n with a leak rate wr dtional assuranee that this grad be datactad and the plant shut dew Gwill m niterien TLEAKACE-preeursar-in a timely manner.rhe limit of 150 arallons per day per SG is based on the o operational LEAKAGE perfoance criernin in NEt 97-06. "Steam Generator Prora Guidelines". The Steam Generator Program operational LEAKAGEoperformance criterion in NET 97-06 states. "The RCS operational rmv to secondary leakage throuah any one SG shall be limited to 150 arallons per day," The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakaae rate criterion in coniunction with-the impl-ementation of the Steam Generator Proaram -is an effective measure for minimizin h frequency of steam _enerator tube ruptures.

BEAVER VALLEY - UNIT 1 B 3/4 4-3g Amen..int.LChanae No. £4-9-029

REACTOR COOLANT SYSTEM Provided for Information Only.

BASES 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

a. Unidentified LEAKAGE7-_oX identified LEAKAGE, or Primary to s...ndary LEAKA.E 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 LEAKAGE, identified LEAKACE, erlril r to seeondary LEAKAE cannot be reducod to within limfits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reacter must be brought to !ewer pressure conditions to reduee the severity of the LEAFCE and its -

p**tntial .ns..u.n..s. The ree*-tr must be brought to MO-DE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reducos the LEAKAGE.

The allewed Completien Times are reasonable, based o~n eperating experienee, to reaeh the required plant eenditions from full power conditions in an orderly manner and without challenging plant systeffs. In MODE 5, the pressure stresses aeting on the RC-PB are much lower, and further VLtErUratIen +s 4-3 mu/ . ss lie.ly.

BEAVER VALLEY - UNIT 1 B 3/4 4-3h Amnc~nemnt-Cb gg~==N. &I0M.

1, -14 REACTOR COOLANT SYSTEM Providedfor Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS (Continued) ab. If any pressure boundary LEAKAGE exists or primar_ 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 3-6-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 leakaae 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 leakaae detection instrumentation reguired 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 reguired 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. Primar. ... y te ndary LEAKAGE is al.. measured by perfermanzc of an RC -

water inventory balarz- in conjunction with effluent menitering within the zcczrndary stea-m and feedwater systems.

Providedfor Information Only.

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

Therefore, this SR is net required to be perfermeel in IWODES 3 and-4 until 12 heurs ef steady state eperatizr. near eperating pressure have been zstablished.The SR is modified by two notes. Note 2 states that this SR is not reauired to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishina 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 menmenC No. 841-29 i

REACTOR COOLANT SYSTEM rSPr~ovidd Provirdde d for Information Only.

. Ony.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

SURVEILLANCE REQUIREMENTS (SR) (Continued)

SR, 4. A . C . t 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.

An early warning ef pressure beundary LEAKAGE er unidentified 13EAAG2E is Previded by the systems that meniter the eentainxncnt zitmesphere radioactivity and the. .ntai*..ntsufp level. The 12 heur m.nit.ring ef the lealeage deteetion system is suffieient te previde an early warning Cf inereased flCZ LEAKAE. These leakage deteetien ytm are speeiified in LCO 3.4.6.1, "Leakage Detectien instrthraentatien."

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. Nate (1) states that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillancc is required enly en leakage deteetien instrumentatien required by LCD 3.4.6.1. This Nete allews the 12 heur menitering te be suspended en leakage deteetiefl instrumentatien whieh is inaperable Cr net required te be epcrable per LCD) 3.4.6.1. Nete (2) states that this SR is required te be perfermcd during steady state eperatien.

Note 3 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 aallons 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 than or eaual to 150 aallons per day throuah any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE nerformance criterion in the Steam Generator Proaram 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 (25 0 C) 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 SRC 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 reauired 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

I 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 usina continuous process radiation monitors or radiochemical arab 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 4-3i AmendmeEEBChanae No. 41-IQ-029 I

Attachment C-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:

Page 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 4-4e*

B 3/4 4-4f B 3/4 4-4g 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 GENERATORS (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. 2-4-22-aD-U I

REACTOR COOLANT SYSTEM Prvded 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.

Ornz ePERUABLE steafa generzater in ai nen iselated reaeter eeelant leep previeles suffieient heat rezmcval zcapability te remeve deeay heat after a rcxazter shutdown. The ~~irmn fer twe O)PERABLE steam generaters, eembirncd with ether irceuiremznts ef the Limiting Ccnditiens fer Gperatien ensures zideeqate BEAVER VALLEY - UNIT 2 B 3/4 4-2 Change No. 2-42---r,031 I

$I Providedfor Information Only.

REACTOR COOLANT SYSTEM Proposedchanges to draftpagefrom Unit 2 LAR 173 (EPU) e'~.r.

docay heat romoval capabilities for RCZ temperatures greater than 359 0F if eno steam generater bocomos- inpera-ble due to single failur considoratiens. Belew 3502F, docay heat iis romovod by the fR4 system.

The Surveillanee Reefirements for inspoctien of the steam generater tubes ensure that the structural integrity of this portion of the RCZ will be maintained. The program for insorvicoe inspectien of stoa generater tubes is based ~n. a modifieation of flogulatery Cuide 1.83, Rovisien 1. inservico inspeetior. of steam generator tu-bing is essential in erder to maintain survoillanco of the eonditions of th tubes in the event that there is ovidenco of mechanical damage or prgrssve degradatien due to design , manufacturi'ng errers, or-inservico conditions that lead to corrosion. insorvic-- nspoctien of steam generater tubing al-so provides a moa;i-ns f- char-acterizing the nature and causo of any tube dogradation se that corrocetive measures ean be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within these parameter- limit found to result in negligible corrosion of the steam gonorater tubes.

if the socondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress rro -- cracking.

ior The extent of cracking during plant oporatior.

wouild be limited by the limitation of steam gonorator tube leakage between the Primary Coolant System and the Secondary Coolant Syst-em (primary to secondary LEAKAE - 150 gallons per day per steam genorator) . Axial crackes having a primary to secondary LEAYACE lossa than this limit dutr-ing operation will: have an adequate margin of-safety to withstand the loads imposed during normal oporation andl byZ postulated accidonts. Operating plants have demonstratod ta primary to socondary LEAKAGE of 150 gallons per day per steam gonorator can readily be doetected. LEAKAGE in excess of this limitt

%All roqfuiro plant shutdown and an unsehodulod inspeetion, during which the leaking tubes will be locatod and plugged or repaired by slooving. The tochieial bases for slooving are described in the approved vender reports listed in Survoillanco Roquiroment 4.4.5.4.a.9, as supplemented by Wostinghouso letter FENGC 02 304.

Wastage type defects are uinlikoly with the all volatilo treatment (AVT-) of socondary coolant. llowover, even if a dofoct of simila-r t:ýT should dovolop in sorvico, it will: be found during schodulod inse-rvico steam generator tube oxeaminations. Plugging or repair will be required of all tubes with imporfoctions oxcooding the plugging-o repair limit. Degraded steam gonorator tubes may be repaired b~yth installation of Sleeves which span the degraded tube seeton A steam gonorator tube with a sleeve installed meets the structural BEAVER VALLEY - UNIT 2 B 3/4 4-3 Change No. 2 GIG2--D2m I

II I REACTOR COOLANT SYSTEM BASES I Providedfor Information Only.

3ý4.4.5 ScPE;AM GEwE;iýAT-GRS (Gent+nue

-,gIrama-- nts of tubes whieh are net degraded, tharafere, the sleeve is ..nsid.red a part ef the tube. The survvillan- requirements identify these slacying mathedologias approvad fer use. !If an19 installed sicaeve is feundl to have threugh wall panaetratien greaater than or equal to the plugging limnit, the tube must be plugged. The plugging limit for the sleeve is daerived from Rl. C. 1.121 analysis whieh utilizes a 20 pareent allowanca for addly eurrant uneartainty i~n determining the depth of tube wall penetration and ad*ditional d.gradati.n growth. Steam generator tube inspeetiens of eparating Plants have damonstratad the eapability to reliably datatee dagradlatien that has penetrated 20 pareent of the original tube wall thicecness-.

The voltaga basadraai limits of these survaillanca re ireamants (SR1) implement the gu-idlanea in GL 95 05 and are appliaabla only to Westinghouse designed steam ganarators (S~s) with outsidae diameter-stress aorrosion aracking (ODSCC) locatad at the tuba to tuba support plate intarsaations. The guidanca in GL 95 05 will not be applied to the tuba to flow distribution baffle plate intarsaatioens. The voltaga based repair limits are not applicabla to ether ferms of SC tuba daegradation nor are they appliaabla to OPSGC that oacur-s at ether locations wi~thin the SC. Additionally, the repair critaria apply only to indications where the dagradation fnachan~ism is domainantly axial GDSCCG wi~th no NDE datactabla cracks axtandling utsidhe the thicer ss

.. of the support plate...r. " to GL 95 05 fo-r additional dascription of the dagradation morphology.

implamantation of these Sils requires a darivation of the voltaga structural limit from the burst versus voltaa ampir. ical corrlation and than the subs...- nt darivation of the.. taga repair lim44it from surveillance).

the ' ...e stru.. ....

sral lii e is the .. *..~

imleene _*. by %this.

The voltaga structural limnit is the voltaga fromf the burst prassura/bobbin voltaga corralation, at the 95 parcant pradiction interval curva raducad to account for the !ewer 95,195pran tolar-anca bound for tubing material: preper-tias at 650 0F (iZa., tha 95 parcant LTL1 curva). The voltaga structuiral limit must be adjusted dewnwardl to account for potantial daegradation growthduiga operating interval and to account; for ?ME uncarctainty. Th upr voltaga repair limit; VURL is detemined from the structuralvotg limit by applying t-ha folowng abaien; BEAVER VALLEY - UNIT 2 B 3/4 4-3a Amen tCha No-.-231 No. I

1~

II I REACTOR COOLANT SYSTEM BASES I Providedfor Information Only.

Proposedchanges to draftpagefrom Unit 2 LAR 173 (EPU) wher'e-V~ raprýsants the allewanee f-eir degradation grawth between inspactians and VtID represents the allawanea feLr petential: seur-ees ef eLrrar in- t-h-lmeasura-ment af the bebbin ccil veltaga. Further diseussien af the assumptiens nacassary te dateLrmine the velta-ge repair limit are discuss. d in GL 95 05.

Safety analyses wr,* parf-rmed pursuant to Cenrie Letter 95 05t deter-mine the Fmaximum MSZLB indueed primary te seeandary icale rate that eeauld eecur- withett effsite daeses emeding a small: fraetian af 10 CFR 50.67 guidelines (eansidering a eeneurrent iedine spike), 10 GFR 50.67 (pre accidant iedine spilce), and witheut eentrel: raeefm deses axacading 1:0 cmR 50.67. The curraent valuce of the maximum 1MZLB induced leak L-at- and a summary ef the analyses area pr.vidd in Seetiefl 15.1.5 Cf the UFSAR.

The m.id .yel e,.quation in SR 4.4.5.4.a.1g.d sha.ld. only -e used far indieatiens at the tube suppert plates.

SR 4.4.5.5 implements several reperting r giaats racCemmnded by GL 95 05 far situatians which tha NprC t ta be natifiad prier te raeturtning the S~s ta sarvica. Far- the pur-pasas af this raeparting ragirmat, icakeage and canditianal burst prabability can be calclatd based en the as fauind valtage distribution rather than the prajacted and af cycle (EOC) valtage distributian (refer tC GL 95 05 far more infarmatian) when it is net practical to complete these

ý *== i 4 4c=- -- 4- =- n ~ m J-= - A4-J = ]*ý"* 1 - -q. - - i-=

ratur-ning the S~s ta sar-viea. Nata that if lcakeaga and eanditianal burst prabatbility were ealetulatad uising the maasuraed EEOC valtaga distr-ibutian far- the purpasas af addressing the GCL sactien 6.a.1 and 6.a.3 rapeirting criteria, than the results af the praejectad EGG voltage distribution shauild be provided par the GL sactian 6.b (a) eritaria.

W~henever the raesults af any steam ganaratar tubinginrva inspeetian fall inta Categary C 3, these raesults will be repar-tad ta the Cafmissien pursuiant ta Spacifieatian 6.6 prier ta rasumptian afE plant aparation. Stuch easaes will be considarad by the Cammissina a ease by ease basis and may raesuit in a raia tfar- analysis, labar-atery examinatians, tests, additianal eddy curraent inspactian, and.

re is.e ef .i'.

therN. lSe==a. Spe Stt~ a tin 1 .J~. a it: . neee s sar-3/4.4.5 Steam Generator (SG) Tube Intearitv 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 aenerator tubes are an integral part of the reactor

lI W IProvidedfor Information Only.

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).

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

Specification 6.19. "Steam Generator (SG) Program." reguires 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 aenerator tube rupture (SGTR) accident is the limiting desian 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 eaual to the operational LEAKAGE rate limits in LCO 3.4.6.2.c. "RCS Operational LEAKAGE." plus the leakaae 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 discharc dthrough the main condenser.

For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EOUIVALENT 1-131 is assumed to be egual to the LCO 3.4.8. "RCS Specific Activity," limits. Pre-accident and concurrent iodine spikes are assumed in accordance with applicable reaulatory auidance. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaaed fuel. The dose conseauences of these

) a IProvidedfor Information Only.

events are within the limits of 10 CFR 50.67 as supplemented by Reaulatory Guide 1.183 and within GDC-19 values.

The analysis for design basis accidents and transients other than a SGTR assume the SQ tubes retain their structural integrity (i.e..

they are assumed not to rupture) and the steam discharge to the atmosphere is assumed to include primary to secondary SG tube LEAKAGE ecruivalent to the operational leakaae limit of 150 agd oer SG.

However, an increased leakage assumption is applied in the Unit 2 MSLB analysis. In support of voltage based repair criteria pursuant to Generic Letter 95-05, analyses were oerformed to determine the maximum permissible main steam line break (MSLB) primary to secondary leak rate that could occur without offsite doses exceeding the limits of 10 CFR 50.67 as supplemented by Reaulatory Guide 1.183 and without control room doses exceeding GDC-19. An additional 2.1 apm leakaae is assumed in the Unit 2 MSLB analysis resulting from accident conditions. Therefore, in the MSLB analysis, the steam discharge to the atmosphere includes primary to secondary SG tube LEAKAGE eguivalent to the operational leakaae limit of 150 god per SG and an additional 2.1 gpm which results in a total assumed accident induced leakaae of 2.4 gpm.

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

LCQ The LCO regquires that SG tube integrity be maintained. The LCO also reguires that all SG tubes that satisfy the repair criteria be pluaaed or repaired in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Proaram repair criteria is repaired or removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not oluaaed or repaired, the tube may still retain tube In the context of this Specification, a SG tube is defined as the entire lenath 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 oerformance criteria. The Sr 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 oerformance 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.

i Providedfor Information Only.

Tube burst is defined as. "The aross structural failure of the tube wall. The condition typically corresponds to an unstable openina 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 intearity performance criterion provides guidance on assessing loads that have a sianificant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loadina 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 reguires 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 desian basis loads based on ASME Code. Section III.

Subsection NB and Draft Regulatory Guide 1.121. "Basis for Plugging

Dearaded Steam Generator Tubes,

" August 1976.

The accident induced leakaae 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 as described in the Applicable Safety Analyses section. 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 durina olant operation. The limit on operational LEAKAGE is contained in LCO 3.4.6.2. "RCS Operational LEAKAGE." and limits primary to secondary LEAKAGE throuah any one SG to 150 aallons per day. This limit is based on the assumption that a single crack leakina this amount would not oropaaate 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 c~nm41 1 An A i-'h= ;4hn'rcn7- nc!ýiwwi -in n 1 i'f -xc 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.

SProvided for Information Only.

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 regquired actions provide appropriate compensatory actions for each affected SG tube. Complyina with the required actions may allow for continued operation. and subseauently affected SG tubes are aoverned by subseauent condition entry and application of associated reguired 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 pluaaed or repaired in accordance with the Steam Generator Proaram as required by SR 4.4.5.1. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam aenerator tube intearity 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 arowth 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 pluaaed 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 outaae or SP 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 arowth 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 pluaaed or repaired prior to entering MODE 4 following the next refueling outaae 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 hQur-s

.1 I Providedfor Information Only.

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 challenaing plant systems.

SURVEILLANCE REOUTREMENTS SRA4.45-1 During shutdown periods the SGs are inspected as reniired by this SR and the Steam Generator Program. NEI 97-06. "Steam Generator Proaram Guidelines" and its referenced EPRI Guidelines establish the content of the Steam Generator Program. Use of the Steam Generator Proaram 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 ourpose 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 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 existina 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 Proaram defines the Frequency of SR 4.4.5.1. The Frequencv 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 arowth rates to determine an inspection Frequencv that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition. Specification 6.19 contains urescriptive requirements concernina 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 satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw arowth. 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).

Only.

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

Steam aenerator 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 pluaged or repaired prior to subjecting the Sr tubes to sianificant primary to secondary pressure differential.

BEAVER VALLEY - UNIT 2 B 3/4 4-3b Change No. 2--0-1-2-03 I

REACTOR COOLANT SYSTEM R 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 2 B 3/4 4-4d Amendment No. 64

REACTOR COOLANT SYSTEM ProvidedforReadability 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 150 gpd per steam generator 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 Change No. 2-034

I REACTOR COOLANT SYSTEM Providedfor Information Only.

Proposeddraft page froni Unit 2 LAR BASES 1- 173 (EPU) 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

APPLICABLE SAFETY ANALYSES (Continued)

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.

Due to adoption of the voltage based steam generator tube repair criteria per guidance provided by Generic Letter 95-05, the safety analysis for an event resulting in steam discharge to the atmosphere conservatively assumes a 450 gpd primary-to-secondary LEAKAGE (150 gpd per steam generator) for all accidents other that--!lbathe MSLB.

The dose consequences associated with the MSLB addresses an aeeident-ndiueed-additional 2.1 apm leakage, which, per Generic Letter 95-05, is postulated to occur (via pre-existing tube defects) as a result of the rapid depressurization of the secondary side due to the MLSB, and the consequent high differential pressure across the faulted steam generator. The maximum allowed *l!accident induced leakage is 2.1 2_J4-gpm.

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 Change No. 2 Ola2-M

I REACTOR COOLANT SYSTEM I Provided for 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 I Operating empericnee at PWR plants has shewn that sudden inercascs in lcale rate are eften preeursers tc larger tube failurces. Maintaining an eperating LEAKAGCE limfit ef 150 gpd per steam generater will minimaizc the petcntial fer alarge L3EAKAGE event at pewcr. TPhis epcrating LEA4IACE limit is mere restrictive than the 'prating LEAKAGE lim in stan~dardized tehni.al sp..ifieatitns. This prvides additienal margin t ,

ain cntdat a itubc flaw whi-h might grew at a greater than expecctecd rate er ene eetedly exten~d etutsidc the thickness of the tube suppert plate. This mcnitring rcg ----- *vdes adclitional assuranee that this preeursc EKC will be deteeteel and the plant shut deif~q in a timlely mann.r.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 _aallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakaae rate criterion in coniunction with the implementation of the Steam Generator Proaram is an effective measure for minimizing the freauency of steam aenerator tube ruptures.

BEAVER VALLEY - UNIT 2 B 3/4 4-4g Amenement-Qange No. 1--8-2-031 I

.REACTOR COOLANT SYSTEM Provided for 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 AmnementChanae No. 1O"2r-M31 I

REACTOR COOLANT SYSTEM SEAO C N Providedfor Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS bA. Unidentified LEAKAGE7-or identified LEAKAGE, or primary t.

se..ndary LEAI.C. 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 LEBAKACE, identified LEAKACE, orprmr to sieendary 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 LEAKACE and its potential consequences. The rsauter must be brEught the Mea 3 within 6 hburs and MODE 5 withinn 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This aetien reduAEs the LEAKAscE.

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.

ab. If any pressure boundary LEAKAGE exists o~r Dprimarv to secondary LEAKAGE is not within limit, or if unidentLified or identified LEAKAGE cannot be- reduced to within-limit-a within 4--hours, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals 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 G-6--Jthefolwig 0hours. 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 Ahanae No. -2-02M31 I

'II REACTOR COOLANT SYSTEM Providedfor Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

SURVEILLANCE REOUIREMENTS (SR)

SR 4.4.6.2,A An early warning of oressure 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 leakaae 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 leakaae detection instrumentation reguired 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 monitorina to be suspended on leakage detection instrumentation which is inoperable or not reguired 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. ndary LEAKAGE is a... measured by perfermanee Cf an RCZ water inventory balance in eenjunetier. with effluent menitering within the seeendary steam and feedwater systemsz.

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

Theref ere, this SR is net required te be perfzrmed in MODES 3 and 4 until 12 heurs ef steady state eperation near Cperating pressure have been established.The SR is modified by two notes. Note 2 states that this SR is not regquired 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.

An early warning ef pressure beundary LEAKACE er unidentified L3EAKACE is previded by the systems that mCniter the eentainment atmesphere radieactivity and the eentainment stunp level. The 12 heur menitering Cf the loaleage deteetien system is suffieient tC previde an earlyJ

r Providedfor Information Only.

warning of inereased RGS LEAKAGE. -These lealeage deteetien systems are speeified in LCO 3.4.6.1, "Lealeage Deteetien instrtunentatien."

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 surveillanee is required only en leakage deteetion instrufnentatien required by LCO 3.4.6.1. This Nete allows the 1:2 heur menitering to be suspended en leakage detection instrumentatien which is inoperable or net required to be operable per LCO 3.4.6.1. Nete (2) .tatesthat this SR is reequireed to be Performed during steady state eperatien.

Note 3 states that this SR is not applicable to r~rimarv toseodr LEAKAGE because LEAKAGE of 150 aallons -per day cannot be-maue accurately by an RCS water inventory balance.

SR 4.4.6.2.c This SR verifies that primary to secondary_ LEAKAGE is less than or equal to 150 aallons 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 aallons per day limit is measured at room temperature (25'C) 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 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 Frecuency 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 arab sampling in accordance with EPRI.

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

BEAVER VALLEY - UNIT 2 B 3/4 4-4j Ar ..... .. Chan-e No. ---D- I

v t e Letter Sequence Supplement
TAC:MC8861, Steam Generator Tube Integrity (Approved, Closed) TAC:MC8862, 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-06-119, Supplement to License Amendment Request Nos. 324 and 196, Steam Generator Tube Integrity

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L-06-119, Supplement to License Amendment Request Nos. 324 and 196, Steam Generator Tube Integrity

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