Text

Supplement to License Amendment Request Nos. 324 & 196 Steam Generator Tube Integrity
	ML061580610
Person / Time
Site:	Beaver Valley
Issue date:	06/01/2006
From:	Lash J; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-06-088, TAC MC8862, TAC MC8861
	Download: ML061580610 (94)
	v • d • e

FENOC FirstEnergyNuclear OperatingCompany James H. Lash 724-682-5234 Site Vice President Fax: 724-643-8069 June 1, 2006 L-06-088 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 November 7, 2005 (L-05-144), the FirstEnergy Nuclear Operating Company (FENOC) submitted 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 letter dated March 1, 2006, the NRC requested further information regarding the FENOC submittals. In a letter dated April 25, 2006 (L-06-063), FENOC provided a response to the NRC request with a commitment to provide supplements to the LAR Nos. 324 and 196 that would incorporate modifications proposed in the response. Attachments A-1 and A-2 are proposed BVPS-1 and BVPS-2 Technical Specification (TS) changes. Attachments B-1 and B-2 are proposed BVPS-1 and BVPS-2 TS Bases changes. The proposed TS Bases changes are provided for information only. All of these attachments are provided to replace the corresponding attachments contained in the initial LAR submittal. These attachments contain revised markups reflecting the following:

1. Changes needed to reflect responses to the Request for Additional Information (RAI) dated March 1, 2006. Affected attachments have been annotated by identifying the corresponding RAI item number in the page margin adjacent to the area containing the associated revised markup. The revision is highlighted to distinguish it from the original markup.

2. Changes required for consistency with BVPS-2 LAR 173 (Extended Power Uprate). These changes are provided because issuance of a license amendment for LAR 173 is now expected to occur prior to approval of BVPS-2 LAR 196, rather than after as assumed when LAR 196 was submitted.

Affected attachments have been annotated with "EPU" in the page margin adjacent to the area containing the associated revised markup. The revision is

Beaver Valley Power Station, Unit Nos. 1 and 2 Supplement to License Amendment Request Nos. 324 and 196 L-06-088 Page 2 highlighted to distinguish it from the original markup. These changes would involve direct incorporation of information expected to be reviewed and approved by the NRC through LAR 173. Therefore, additional technical review of these items (e.g. revised tube repair sleeve repair criteria and revised accident induced leak rate values) should not be required.

To ensure consistency between the license amendment for LAR 196 and other pending license amendments, it is requested that the amendment for LARs 324 and 196 would be approved by August 1, 2006, to be implemented for both units prior to the first entry into Mode 4 during plant startup from the 2R12 refueling outage planned for the fall of 2006.

Attachment C provides a tabulation of miscellaneous changes that (1) are not related to the RAI or incorporation of EPU, or (2) are related to the RAI, but differ from the RAI response. These changes are also highlighted in the affected attachments, but not annotated.

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 commitments are contained in this submittal. If you have questions or require additional information, please contact Mr. Gregory A. Dunn, Manager, Fleet Licensing at 330-315-7243.

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

Sincerely,

. es H. Lash Attachments:

A-1 Proposed BVPS-1 Technical Specification Changes A-2 Proposed BVPS-2 Technical Specification Changes B-1 Proposed BVPS-1 Technical Specification Bases Changes B-2 Proposed BVPS-2 Technical Specification Bases Changes C Miscellaneous Changes

Beaver Valley Power Station, Unit Nos. 1 and 2 Supplement to License Amendment Request Nos. 324 and 196 L-06-088 Page 3 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-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Changes License Amendment Request No. 324 The following is a list of the affected pages:

Page V

xv 1-3**

1-4 3/4 4-8 3/4 4-9 3/44-10 3/4 4-10a 3/4 4-10b 3/4 4-10c 3/4 4-10d 3/4 4-1Oe 3/4 4-13 3/4 4-14 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 3/4.4.6 STEAM GENERATORS (SG)

REACTOR COOLANT SYSTEM LEAKAGE Tube Intearitv....... 3/4 4-8 I

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. 2-" 1

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

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

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

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

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

LEAKAGE 1.14 LEAKAGE shall be:

a. Identified LEAKAGE

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

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

DEFINITIONS

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

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

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

1.15 THROUGH 1.17 (DELETED)

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

DOSE EQUIVALENT 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-131D E = cI- 13 1 + CI- 1 3 2 + CI-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. 24-6 1

ft+/-!dAUTUL tU(l.~dT1 b~, +/-!

3ý44. A STFaa E Nm~aff7GmsT LI'31ITINC CONEqD;ITION r. . OPER-T 3.4.5 Eaeh steam generator shall be OPERAB-LE.

APPLICA!ILITY: WD.ES 1, 2, 3 and 4.

With .n. .r m.r. steam generators in"p.rable, r..t-r- the inoperable generater(s) to OPERABLE states prier to nraegT above 200'r-.

SUVILLANGE REQUIREMENTS 4.4.5.1 Stcam C.n...ater "-%`^ Scctien and^ In..p..ti B.eThe a

steam generater shall be determined PE LE during shutd.. .... b seleeting and inspocting at least the iiu ntuber of steam.

generaters "peified in Table 4.4 1.

4.4.5.2 Steam ConeratorF Tube S-ame-le Seletion and insneetiR The steam g.n.rator tu4b miimmple size, -nspeetlen result

.las.ifieation, and the .. rrsp..ing" a-tion -,- red shall be as speeified in Table 4.4 2. The inserriee Inspeetien of stea genorvator tu-bes shall be perfermed at the frequeneies specifiod In Spocifieation 4.4.5.3 and the inspeeted tubes shall be verified aeceptable per the aecopt-anee criteria of Speeifieation 4.4.5.4-Steam gonerator tu-bes shall: be examined in aecordanco with Articlo of Seeti-n V ("Eddy .ur*r.. nt. E-.amination of Tubular

.r.duts^") and Appendix IV to Seetion X1 ("Eddy Current Examination of Honforromagnetie Steam Coner-ator Heoat Emehanger Tubing") ofth applieable year and addenda of the AZME Doiler and Pressure Vessel Code required by 10CFR5O, Section 50.55a(g). The tubes selected for each inservico inspeetion shall include at least 3 percont of tho total n'ber of tubes in all steam generat.r.; the tubes seleeted for these inspeetions shall be seleeted on a randomf baaia xot

a. WThere expetienee in similar- plants with similar water ehomistry indicatoa erieatal areas to be inspcted, thcn at least 50 percont of the tubes inspected shall'be from thesac critical ar*as.

b. The first sample of tuboc selected for eaeh ia~c inspoction (subsequent to the prcovio naotien) of oaech steam genorator shallinluo

1. All nonpluggod tubes that provioucsly had deteetable wall penetrations greater than 20 pereent, and-flP7;UPT? U!'YT T P' T=T-TT 1 m3mn Ne.

Amendment o 7 274

REACTOR COOLANT S*STESM SURVEILLANCE REQUIREM~ENTS (Ccntinuzd)

2. Tubes in these areas where excinee has indieated petential prablzmaft, and

3. A tube inspeetien Pursuant te Speeifieatic 4.4.5.4.a.8 shall be p^rf.rm.d en eaeh seleeted tube.

If any seleeted tube dees net permit the passage z the eddy eurrent prebe fer a tu-be inspeetiern, t shall be ree-rded and an adjaeent t-b shall be seleeted and subjeeted te a tube inspeetien.

e. Theic taube seleeted as the seeend and third samples (i-f r -,I by Tabl -j 4.4.a.h

2) during in--... in-p- tie..

mayr be subjeeted te a partial tube inspeetien previded:

1. The tubes seleeted fer these samples inelude the tubea frcm these areas ef the tube sheet array wheLrz ttbz with imperfeetiens were previeusly fetind, and

2. TPhe inspeetiens ineltide these pert16n-9 Cef the ttubea uml=:Pý =:= I=$=;= Pka-1=6

- --. ~---

BFEPAFER VALTMEY UNITT "r l' 1 DEAVR VALEYAmendmentNe. 27-3

REACTOR COOLANT SZYT-EM SURVEILIANCD REQUIRR5ENTS (Ccntinued)

The rcasults ef eaeh samnple inspeetizn shall be elaseifled intz ene ef the f-llewing

.aterc - three Gateeei=,FiflCDCti' en Resttlt~

C 1: Less than 5 pcrznt ef the t.tal tubes Inspeeted are degraded tibes and nzne e-f the inspeeted tubes are defeetiv.

C 2 One er mere tubes, but net mere than 1 percent -f the t-tal tubes in"p.ctd ar df "tiv., r between 5 p..... and 10 percent ef the total tubes in"peeted are degraded tubes.

C 3 Mere than 10 pereent Cf the tetal tubes-in.pecttd are degraded tubes Cr mere tha 1 pcrccnt Cf th- _in"p-tzed tubes ar defeetivec.

Nete: in all inzipeetiens, pi-evietusly degraded. tubes must e~d-iibi-t signifieant (greater than 10 pereent) further wall1 p-n-tratie-nz3 t- be in"luidd in the abev- pcrccntage calculations.

4.4.5.3 insveetien Freeaicnciczi The abeve reefuired inserviee inspeetiens Cf steam generater tubes shall be perfermed at the fellewing freelucncicC:_

a. The first .

"nr "-

i. in"p..ti.n Cf the Medel 54F st g.n.rater. shall be perfec.,d after 6 Effeetive Full Pew cnth but- within 24 -al-ndar m-nth-

f initial eritieality fellewing steen generater replaeement.

Subscecunt inservicc Inspcctiens shall be p-rferm-d at intzrvals of net . less than 12 n. r.. :than 24 calndar months after the -rvius Inspectin. If twe ccnseetutive inspeetienzi, net including the preserviee Inzspeetlen, rceult In al! inspeet.. n r-.l.. falling int- the C 1" eategery Cr if twe eenseeutive inspeetlens demenstrate that previeuisly ebseL-ved degradatien has net eentinfued and ns additi.nal d.grad.ati.n has o...rred, the inspeetlen interval may be extended te a mama: of enee pecr 4 9 menths-.

Ncte: inserviee in-peetien is net required during the stecam.

aeneratce_ _e . -ement eutacaC.

BEAVER VAIAiSY T iTTlm Vmendment Ne. 273

REACTOR COOLATMý sy-STEM SURVEIL6LANCE REQUIREM1ENTS (Cantinuaed)

b. If the r.stilts ef the in vrvica in.p..tin

" ef a staf generator . -ene,--tadin a. ..rdan. w. ith Table 4.4 2 fall inte Categery C 3, the inapac-tien freefdney shall be increased to at least once per 20 mnths. The increase in inspeetlen freeaeney shall apply until the subaaeant inspactiena satisfy the criteria ef spacificatin 4.4.6.3.a; the interval may then be extended tC a -x 1

-f .. per 4.. . canths.

e. Additienal, tinsheduled it. ------ inspectians shall be parfermad en eaah steam generater in aeeerelanee with the first sample inspactien specified in Table 4.4 2 during thea aIhutctwn subaaqant: te any ef the3 rfllewing eenditiens:

I. Primary tC seeendar- tube leaks (net incluiding leaaka CrIginating frem tube tC tuba sheet welds)inaca afthe limits ef Spac!ifiatien 3.4.6.2-,

ni 'A 0,Uý .4 . S'%.~ . . .~&~.S.. L&t* ,,U ý 4L4 ^J J IA- L.4 .IC O Sarthqfaake-,

3. A less Cf eeelant aeeidant n actuatien ef the en.gi-eard safeguards, Cr aflf M-m MM ; QI=OMV!A4 MIR ~ iflc 4 :* 44=r=

4.4.5.4 Aeeeittanee Critcria

"* 3__ **__'

"', iiQ',= :*in t=Rni' -now i in-

i. imigerfeeti.. means an aircaptian te the dimansien&-,

finish er eanteur ef a tuiba fram that required by fabr..ati.n drawings Cr s.e..fi.atian.... .. y currant testing indicatiens belew 20 pereant ef the rnzmina tube via!l thi'lme". , if d"tc-tabl', may be. .n.idered as imperfectiens.

2. Daczradatien means -a cr-ica inducad crac'ing, wastage, wear er general eerresian Cccurring en aiher inside er eutsida at a tuba.

3. Degraded Tuba mean ^entai.*.n a tuba i"parfecti.n.

greater- than equal

.r t 20 p.rcant af the neminal wall thielnaao causad by dagradatianR.

4. Parcant Daeradatien means the pareantaga ef the tuba wal thickeness affeeted or rameved by degradatien.

PRAIflT? VM4A.T.1RY ZZT7TE 1Imnir~tN MA 7 071

REACTO13R GGGIANT S*STEHz

_ 1.

5. Dofect means an impeirfectien of such severrity that i-t eoeeeeds the plugging limit. A tube containing a defect is defective. Any tube whieh dees net perit the passage of the eddy eurrent inspeetien prebe shall be deemed a defective tube.

6. Plucrcrfinef Limit means the imperfectien depth at or b.y.nd whi"ch the tube shall be removed from sorvico by plugging beeause it may beeome uns.rvicablr.

to the next inspeetien. The plugging limit is9 a to the 40 percont of the n.minal tube wall thickn-4s.

7. Unservicoable describes the conditie.n of a tube if it-Ireaks er cotalins anAdefet large eneugh to affect its structural .integrityIn the evnt - f an Operating Blasis Bar-thefcale, a less of coolant acc-ident, or a steeanline or feedwater line break as specified i 4.4.5.3.e, above.

8. gTabe Ins~oetien means an Inspection of the steo generator tube from the point of entry (hot leg side) eempler-ey aretndth " Ot!ft- __ -IM tý_Qv 5wvUI:e at.

the cold leg.

m ecmpleting the crrosponding a.tions (plug all tub.Sm emeeeding the plugging limit) L-eefired by Table 4.4 2.

a. Within 15 days following the , empl.ti.n of ah - r*. v.i '

inspo,ti. n of steam gne.rat.r tubes, the nv ab.r of tuiba plugged in each steam generator shall be submitted in a Speeial Report in accordanco with 10 GFR 50.4.

b. The complete results of the steam generater tube inservico_

o

1:mQ*! --:1AQ - n- 4 nr Q1nR~Tn1 :rr1- :1n a Spee~a+/- +/--e.per = n *1*

a...rdane. with 10 c- 50.4 within 12 months following th.

eempletien of the inspeetion. This Speeial Report shall-inelude:

nij1mr-wwr ,4noq 03c:mj f tubes inspocted I:%

2. Location and .. r..nt of wall thiknoss tD.n.tratien

.0 U .1 4 .1 c er eae RM e== en " an raper eut en.

2 T.4,Q=I= i FU Omni-_M' A:n Af 4=;4mýPQ :m I ;u*0*04 BEAVER VALLEY UNIT 1 mndotNo Amendment Ne. 2747

RDbACTGR COOLANT S*SgTE~A SURVEILLANCRE REQUIRE1499PPZ (Ccntzinued.)

e. Results ef steam generater tuibe inzspeetiens whieh fall inte Categery C 3 shall be repeirted te the Gciac purstuant te Speelfieatien 6.6 pr-ier te irzsufqptien ef pl!n eperatien. The written repeLrt shall previde a eleseriptin

.,f investigatlens eendueteel te deter-mine the eauzsz ef th tube degradatien and eerreetive measures takcen te prcevent-reeturrenee.

BFA-VER VALLEY 3ý4 4 lGe BEAVER~~~~~~A m~~2t h 73:

PALY.:4-Q

SABLE 4.4-1 A--

,L MINIMUM MURIBER STEAM GENEA.*L.')R* TO* E1 INSPECTED DURING!C N.E.VIC- INSPECTIO-N Prrcscrviee iflsp-ebti~n Ne______________

Ne. ef Steam Gene-raters per-Unit- q~iee que

-I'rst Inserviee Ifqpcin______________________________

Table Netatiznf.j (1) The inzv . e inspeeetizn may -be limited te ene steam generater en a rzetating sehedule enc9mpa aong 9 p-re-nt - f the tubez If the rcsults ef the first er pr------- 'inspeeti indcat-z-#-enthl-,,at all steam or n'ratftor are p'rfc'mn".i......n a lil--mannr. N.t. that tm...r

- .rcutmstanczzi, the eperrating czrnditiens in ene er mzrz steam generaters may be feund tcs-me b

... r. severe than thezs in ether steam gtneratcrs. Under such eircumstanees the sampl sefuencz s-hall be ... dified te in.p..t the m..t severe eenditiens.

(2) The ether siteam grcnerater net iniiezcted edurincg the first inszric izizticn shall be insDocted. The third and subsc*-uent ins;--tiens sheuld fellew . .nstruebien-the .. serib..in S'~P~T r p~7 ~ I ;7t4~4~4Q4 Amcncimznt zlc. ~;i 3/4 4 l0d i ZH6

TABLE 4. 4-2 STEM1 G&N'ERA-TOR TUBE INSZPCTO Btt R

w~nere n ais rar.t zr efrnz eam czcac3a~cziarnra nrzan I3EAýfR VA~LLEY UN~q3IT -- 4 O 3ý4 ~zdnn 7 Amenelment Ne. 2q3

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 SG tube integrity shall be maintained 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.

- --- 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 oluaaed in accordance with the Steam Generator Program:

I. Verify within 7 days that tube intearity of the affected tube(s) is maintained until the next refueling outaae or SG tube insoection.

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

b. With Action a not being comnleted within the specified completion time or if SC 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 IL SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

.4-5.1 Verify SG tube integrity in REQUIREMENTS accordance with..and at a t-hp qt--P;;m rpnprqt-nr Prncyrqm-SURVEILLANCE 4.4.5.2 Verify that each inspected SG tube that satisfies the tube renair criteria is pluaaed in accordance with the Steam Generator Proaramnrior to entering MODE 4 following a SG tube inspection.

BEAVER VALLEY - UNIT 1 3/4 - Amendment--NQ--

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:

bg. With any Reactor Coolant System operationalLEAKAGE greater than any nxylfdifngtfor ene ef t abeve-... *Dlimits, 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 Ho Z STA Y within th next .D6 heursou sand . in. COLD s .. S!;T-DGWN . .... within the fellewir.g3 ek. With the reauired action and associated completion time of Action a not met" or with aey-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-ol 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 1 3/4 4-13 Amendment No. 44& 1

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 )Fast once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state eperatien.

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

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

[3) Not applicable to nrimary to secondary LEAKAGE.

BEAVER VALLEY - UNIT 1 3/4 4-14 Amendment No. 1-&-3, 1

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

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

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

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

6.9.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection 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 dearadation (as defined in EPRI.

ziuri* '*ugiz**z*

.tr

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 pluaced during the inspection outaae for each active degradation mechanism.

f. Total number and nercentace of tubes plugged to date.

a. The results of condition monitorina, includina the results of tube Dulls and in-situ testing, and

h. The effective plugging percentage for all plugaing 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. 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) Proaram 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 followina 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 leakaae. The "as found' condition refers to the BEAVER VALLEY - UNIT 1 6-26 Amendment No. 2-3-4 1

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

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

accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam aenerator tubes shall retain structural integrity over the full ranae of normal operatina conditions (including startup, operation in the power ranae. hot standby, and cool down and all anticipated transients included in the design specification) and desian basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 rgainst burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents.

or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute sianificantly 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 leakaae rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakaae 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 thg eige!Mtional "leakaa jexcent during a SG tube rupue

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

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

BEAVER VALLEY - UNIT 1 Amendment No.

ADMINISTRATIVE CONTROL-S Steam Generator Proaram (Continued)

d. Provisions For SG Tube Inspections Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detectina flaws of any tvpe (e.a.. volumetric flaws, axial and circumferential cracks) that may be present along the lenath of the tube. from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair Criteria. The tube-to-tubesheet weld is not part of the tlube. In addition to meeting the reauirements of d.1, d.2.

and d.3 below, the inspection scone, insnection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

A dearadation 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 I. Inspect 100% of the tubes in each SG during the first refuelina outaae following SQ replacement.

2. Inspect 100% of the tubes at searuential periods of 144. 108. 72. and. thereafter. 60 effective full power months. The first seauential period shall be considered to begin after the first inservice inspection of the SGs. In iadditicn. During e*ach period inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three interval8s 'between refuelin*

outaaes (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 dearadation mechanism that caused the crack indication shall not exceed 24 effective full power months or one interval beween refueling outages (whichever is less).--If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or encineerina 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 BEAVER VALLEY - UNIT 1 6-28 Amendment No.

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

New page

INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3.3.5 Remote Shutdown Instrumentation ............... 3/4 3-52 3/4.3.3.8 Accident Monitoring Instrumentation .......... 3/4 3-57 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4.4.1.1 Normal Operation ............................. 3/4 4-1 3/4.4.1.2 Hot Standby ................................... 3/4 4-2 3/4.4.1.3 Shutdown ..................................... 3/4 4-3 3/4.4.1.4 .1 Loop Isolation Valves - Operating ............ 3/4 4-5 3/4.4.1.5 Isolated Loop Startup ........................ 3/4 4-6 3/4.4.3 SAFETY VALVES ................................ 3/4 4-9 3/4.4.4 PRESSURIZER .................................. 3/4 4-10 3/4.4.5 STEAM GENERATORS (SG) Tube Integritv.......... 3/4 4-11 I 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 Leakage Detection Instrumentation ............ 3/4 4-17 3/4.4.6.2 Operational Leakage .......................... 3/4 4-19 3/4.4.6.3 Pressure Isolation Valves .................... 3/4 4-21 3/4.4.8 SPECIFIC ACTIVITY ............................ 3/4 4-27 3/4.4.9 PRESSURE/TEMPERATURE LIMITS 3/4.4.9.1 Reactor Coolant System ......................... 3/4 4-30 BEAVER VALLEY - UNIT 2 V Amendment No. 4 1 Cerreeted by letter dated ju2.y 11, 2002.-1

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 REPORTINC 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 .qtem (- enerator Tube Tnsnection Renort .. 6-22 1-ZZ 6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM .................... 6-22g I BEAVER VALLEY - UNIT 2 XIV Amendment No. 1-44 1

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6 - 2 2#

6.12 HIGH RADIATION AREA .............................

6.13 PROCESS CONTROL PROGRAM (PCP) ....................... 6-24 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) ............ 6-25 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and Solid) .................. 6-25 6.17 CONTAINMENT LEAKAGE RATE TESTING PROGRAM ........ 6-25 6.18 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM ......................................... 6-26 6.19 Steam Generator (SG) Proaram .................... 6-27 BEAVER VALLEY - UNIT 2 XV Amendment No. 444 1

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

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

LEAKAGE 1.14 LEAKAGE shall be:

a. Identified LEAKAGE

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

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

3. Reactor Coolant System LEAKAGE through a steam generator to the secondary system (orimary to secondary LI

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 eteem ge**z..rat.r t^be primary to secondary LEAKAGE) through a nonisolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

BEAVER VALLEY - UNIT 2 1-3 Amendment No. 4-"k I

R&GEPCTR COOLDANTR~E Dm A P 3ý4.4.5 sTEAm GENEPATGR-S, T=lMITINC CONDITION FOR OPERATION 3.4.5 Each steam generater shall be GPERALE.

APPLIC**;.ILITY: ... DE. 1, 2, 3 and 41.

With en- er mare steam g.n.rater. ineperable, restere the ineparabla generater(s) te OPERABLE status prier te inereasing TFO"abev&-24G0 F.

4.4.5.1 SteanCnaaa amt 1a --leetien an natien Bech steam gtneratcr chall be determined OPER..LE . ,hutdew. by .. l..ting

.uring and infpicting at least the min.i.....n"... m.r ef -- steam generatr specified in Table 4.4 1.

4.4.5.2 Steam Canerater Tlube S-amele Salactien and nacien The steam generater tube minimtum 3:e

~ampl ize, Inzjpeetien result elazsoifieatien, and the eerrespending actien required shall be as in Table 4.4 2.

.p..ifi.d The in,.m , ,e

.n.pe.ti.n ef steam generator tubes shall be perfemed at the frsnca peeified in Gpeeifieatien 4.4.5.3 and the inspacted tubes shall be verified ae., ptabla per the aeCCptance riteria "f ^p" ifieatien 4.4.5.4. Steia gna*rater tubes shall be examined in aeeerdanee with Artiela 8 ef Seetien V ("E~dd Current Examinati"n ef Tubular Preduets") and Appendix I3V te s.tian X! ("Eddy Currant FExafanatien ef Nenferremagnetic Steam Canarater Heat-Emehangcr Tubing") cf the applicabla year and addenda of the AS B-il-r and Pressure Vessel C.da required by 1bCFy, 0 Section 50.55a(g). Whn applying the e *

'captiens f 4.4.5.2.a thrugh 4.4.5.2.-, prav-ieu defeets er imperfeetions in the &xrca repaired -by slceving are net cansidarad an area r* ... iring rein . p. tien. The tube&

seleeted fer ach. i^ varv **ic*

inspeetien shall inluda at least 43 pereent ef the tetal nu-mbar ef tubes In all steam genaraters; the tubes salected fer these inspeetiens shall be salectad en a randogm basis emcapt:

a. Where exprienee in similar plants with slimilar water eh-mistry indi-at . critical areas te be in^p".tad, then at least 50 parc'nt ' f tha tubes insp- ted shall be fr"m theeo critical areas.

b. The first sample of tubes selectad fer eaah io~c inspeetian (subseefuent to the pr"ric snpactien) ef each steam ganaratar shallinlua TW.ZVlfl 17T.PVR V !TNTUT P 44 11 3ý4

1tklikb.d1U k~iddiid+/- b+/-UUbt+/-!4If%

SURVEIT.Tn~clE; R-BQUI;REBnRN (C-entinueed)

1. All nanpiu~tggad tubes that pravieusly had datactablae wall panetratians greater than 20 p....nt. and

2. Tubes in these areas where axpaianc has indicatead potential problems, and

3. At least 3 peircant ef the tetal nuffber- ef sleeved tubes In all three steaam ganerrater-s. A sample oi1 less than 3 percent is aee...tabl. pr..vid.d all the sleeved tubes in the steam ganar-ater(s) examined during the reftieling eutage are inspacted. These ir..speetiens will ineitude bath the tuibe andl the sleeve, e,14

4. A tube inspactian pursuant te Speeifieatieff

. .4. . any ....... ... .... parmit the passage f the eddy praba fer a.urr.nt tube or sleeve

.is~peetian, this shall: be racardaed and an adjaeeant tube shall be selactad and subjactad te a tub Inigepatian.

S. indicatiean left in sarviea as a racult af appilcatiant af the tuba suippert plate valtaga basad rapair

-ritaria (4..64.a.l(0) shall be Iinopacted by bebbin "i pr"ba " during all futur. rafuaeling outages.

e. The t.ba s. alaatad as the.... nd and third samples (If r------- d y Table 4.4 , j aur-fig eaaan +noarviae Antspacrian may .ubja.tad be t. a partial tuba inop..ti.n pr.vidadn

i. The tubes p alaftid fEr thpese amplo s inelun d the tube&

frem these areas ef the tuiba sheet array wholrae tuba wfith imparfactlicno warae praevieuoly fetund, and-

2. The Inspactians Ineluda these par-tiano af the tuabac wnara Imperractiens wera prcvieauolv feuna.-

I L -- I-

a. +/-mp+/-.m.n.a... ... e th.e steam g.ne.ater tu -- -u-e.oupport plate a-rpai it- r.guiro a 100-parrant rfil bebbin inspactian far- het-lag and eeld laeg tube supper-t plate intaersactiean dawn ta the lawaot eeld lag tuba suppar-t plate with lenawnc eutoilda diameter- strao------ aroin ceracking (ODSCC) indicatianos. The datarminatian af the lewaot eedd lag tuiba suppert plate intaroactiano having GDSCC indicatiaens shall be based efn the parfar-manea ef at leaast a 20 parcaentPal legh r-andam sampling af tubao inopactad avar- thairr 3ý4 4 12 ~44-4-4 ~

Amendmeant Na. 101

REACTOII COOLANT SYSTEM The rcoults ef caeh sample inspeetien' shall be elassificd intz einc e4 the fzllewing threectg~io Lcoo than 5 pc...nt.f thc tztal tubes iraopccted arce deqradcd tubes and nenel ef- the Inspccted tubes are defective.

One or merr tubes, but nrt mcrc than 1 perccnt ef the tetal tube&

in.sp..td a-r defective, er b.tic... 5 pcrcct and 10 p-rcent ef the tetal: tuibes inspcctcd arc degradcd tuibes.

Mcre than 10 percent Cf the total tubes inspeetcd arce Edcgradcd tubes Cr- meLrc than 1: peLrccnt Cf-the in"p..t-d tubes ar" def tiver.

Nete in all inspeetiens, prcvieusly degraded tuabes Cr sleeves must eiehlbit significant (grcater than 10 peLrccnt) frurthcr- wall.

penetr-atleas tC be includcdee in the abeve AA ri,1.,i g4 A~~r -

4.4.5.3 Insigeetien Sraccz . . . bfv. required intrvice a

inspeetieflo Cf steam oenerater tubes shall be perfermed at the fellewing freefuen------

a. The first inserviec inspeetien shall be per-fermed aft;r 6 Eff^"tiv. Full rPewr nths but "alendar within 24 menth ef initial eritieality. GtubsequEnt inserviee Inspeetiens shall be perf rmcd at intervals ef net less than 12 nctr mere than 24 ealendar menths after the .rcviuinspeetien.

if tw" .n....ti. i-'"

"..... 'tinf-ll...."_rvic _ All under V "latile Trcatmcnt 'AP-) ccnditieno, net in"luding the

,rcorvic inspeetien, restult in all inspeetien results falling inte the C 1 eategcry Cr if twe eenoccutivc.

inspeetiens demenstrate that: previecuoly ebserved.

d.gradatien has net eentinued and no additi.nal d*goradatin has eeeturred, the inspeetien interval may be extended tC a maximum Cf enee per 40 menths.

DAVR VALLEY UNh.I1.nmt..

2 3ý4 4 13 Amendment Ne. 1:0

1AcETOR CGOLANTrCYTEN SURVFEILLANCE RLEQUIREMENTS (Ccntinued)

b. !E the iarI ec inspeetien Cf a steam generatzer eendlueted in aeeer-danee with Table 4.4 2 rcquires a third sample inspectien wh"s% :results fall in" .at.g.ryC 3, the inspeetien frceucney shall be inereased te at least enc per 20 m-nths. The in"r.a.. in in"p..ti. n fre-'n-y shall apply ~titl a subse~entz inspeetien demenstrates that a third samuple inspeetien is net rcquircd.

e. Additienal, unseheduled inacr-viee inzspeetizns shall b per-fcrmd en eaeh steam generater in aeeerdanee with the first sample inspeetien speeified in Table 4.4 2 dluring th sbhutdewn subocqucnt te any ef the felleving. c-ditiens.:

1. Primary te seeendary tube leales (net ineluding icalea Criginating frem tube te tutbe sheet welds) in emeesac ofthe limfits ef Speeifieatien 3.4.6.2-,

2. A outrrenee

-1zoi greater than the Operating Basia

3. A less ef eeelant aeeident rciigaetuatlen ef the enginccrzed safeuar-ds, Cr

4. A main steamline er feedwater line bL-arza.

4.4.5.4 Aegeigtanee Criteria

a. As used in this Speeificatien.

1. imeerfeeti-n mzan an ex-eeptien te the dimensien-,-

finish -r eentr Cf a tube Cr sleeve frem that re~ired by fabricatien drawings Cr ccfi Awins.

Eddy eurrent testing indieatiens be13w 20 perezrnt ef the nominal tube wall thiekness, if deteetable, may-be eensidered as imperfeeti4-nS.

2.

Dearadatien means zvc indueed eraeleing,

wastage, wear Cr general eerresien ecurn an either inside Cr eutside Cf a tube Cr sleeve.

3. Dearaded qlbe mctans a tube Cr sleeve eentaining Irncfcin ILItc than er equal te 20 pereent Cf the neminal wall thicknccz eauoced by degradatien.

TRRBPA.MPrf V!A.PAXX TUTNTP P -444 3ý4 4 14 ~

AMf=*qA*Ak=;ql-- h~14 miq i 04-

REACTOR COOLANT SYSTF7M SURVEILL.ANCE REQUIR~iERPZ (Continuoed)

4. Pgreent Docrradatien means the per..ntage ef the tube eLr sleeve well thieleness affected or romevod -by degradat ion.

S. Dfgfct m" means an imporfetien -f uh severity that it oxcoods the plugging or repair limit. A tib corntaining a dofeet is defective. Any tube whieh dec net pemrmit the passage of the eddy eurrent inspeetien prbo shall be deemed a defeetivo tu-boe.

PI6.1t 'Iin-r flopair L.imit. means the imporfoction depth Pl at er- be.end . .hic.h the tube shall be -e r*cd from

.... i.. by plugging or" r."pairo..

d by" l..ving in the affee*td area b..aus. it may b... om.. unserviceable prior to the ictemt irnopeetlio. The plugging or- repai~r limt imperfoction depths are specifiod In porcorntage of nomirnal wall thicknooos as follows.:

a) Original tube wall 40%

This ..

finit.i.. do. not apply to tube au.pp..rt plate irmteLroctions for whieh the voltage based ropair criteria arc being app4i.d. R.f.. to 4..5. .a.0 fo^- the r "pair limit applicablo to these inte-Crooctiens.

b) ADD Coinbustien Engineering TIC welded 3a sleeve wall e) .. laser- wclld sleeve wall

._tingheuse 25-

7. Unsorvicoablc describes the condition of a tube if it leak or nta . a dfct large nough to affoct its structural integrity in the event of an Gp.ratin Basis EarLthqtiako, a loss of coolant aeecidonit, or a steamline or- fodwfator- line brooe as spocifiod in 4.4.5.3.e, above.

0.Tuo npot,.menoa Inspection of the stea~m II - --- -

ee lp4etely .irz-iir th] rr P bond to the top support .'-- R T .f eel-el e BEAVToER VAL3LEY UNIT 2 mn otN.10 Amendment Ne. 1:0-1

SURVEILLANCE R-EQ1JIREMMFPS (Continuoed)

9. Tube Repair roforo te sleeving whieh Is uoode to mait a4i tub0o

- e in Io--V.-cyc r rotuirn a tube to oorvioo. This ineludes the romeval ef plugs that w.r.

inctallod as a ccrroctivo or provontiVe maouro. The fellewing sleeve docsigns have been feund aecoptablo a) ADD Combutistln Engineering TIC weldodl s4love&,-

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

b) W- tirnghei:- laser- welded sl*so.., .83, WAP 1*.

Roviojon 1.

10. Tube Supportz Plate Plugging Limit Is used for- th dioepocition of ant alley 600 steam geneirater tube far continued scrvic" that is oiporioninqg predeminantly amially eriented outsido dliamoeter otrooc corroci erackeing eonfinod wilthin the thickenoco of the tube

.upport plates. At tube tuppert plate interooctions, the plugging (repair) limit is based en maintain-ing steam gonorater tube sorvicoability as deocribd a) Steam gonoratorr tubes, whese dogradateion i attribut to. ..

.utsid diameter stress corroe-i-crackeing within the bouindo of the tube suppor t plate with bobbin veltages less than or equal t 2.0 velts will be allowode to remain In seryico.

b) Steam goneratermp*:4

, 4 in4 3on tubes, 4 whooo deogr-a Eatioen.

attrilbut ed tc out sido diamotorF str-e s e --------

cr-acking within the bouindo ofs the tuabo oupport plate with a bobbin voltago greater- thank 2.0 volts will be repaired or- pluggod, circopt as TRRAIVMR VlxzXT~rhl Am-P4-q A-M-.P:n I-- 14A - I QI

REACTOR O*LANT SYSTEM SURAFEILLANCE REQUIRREENTS (Continued) e) Stczun generater tubes, With indieatiens of potential degradatien attributed to eutsidc dliamcter stress eerresien eLraeling within the betmel f- the tube support plate with a bebbin viltage greater than 2.0 velts but les thn r.

equal to the uppr v.ltag. repair

limitt Ia Loa in in serviee if a roetating pancalee eeil or-

.acceptable alternative in"p..ti. n de-- net d-tcct d-gradatizen. Steam gen.at.r tebs, With indieations of eutside eliamter stresscrro o than the upper voltage repair lmt ilb plugged o VdL.

d) if an Iy+lE

~*nshduld mid in+pr(tion-i pcrfermed, the following mid cycle repair limit apply instead of- the limfits identified in 4.4.5..a.10.a, 4.4.5.4.a.1g.b, and 4.4.5.4.a.!G.e;-

The mid eyele repair limits are determined fromth following eefdatiens.

V

- -2 SL IIA 1.O+NDE+Gr (LL "CL"

_(IT I, )(CL At)

RL- "MRL -URL LRL L i I I

I (1) Thei-ic er- veltaae reiqair limit is ealettlated aeeerdaine to th eth9-'1 sr: in nr T.smtt*A- 'J5 or. Rs~ stmplr e mete BENWER VALLEY 3ý4 4 l4e "A R LEmen4 liN. 101

REACTOR COOLANT SYSTEM SUR-VILL2NCE; REQUIREMEMTZ (Czntinkued*

,.iqt-pp V~l, - '.-pper v- ltage limit VLL~ E 3lweir veltage repair limit

V!Ufl - mid eyele upper veltage repair limit based en time into .y.l^

V1!LI~ mid yele lewer veltage

-rpair limit ba zd en Vmupfl afnd time into eycle At length ef time sifiee last sehe.ulz.

inp...*ztn eluring whi.. ..

implemented CL3 - yele length (the time between twe __h.e,,1z stea" generator V_ = structural limit voltage Cra. average gr.wth rate per cycle length MDE - 95 pereent eumulative pr...b.bility allwan.

fer nondestructive exam.natiln unefrtaint (i.e., a value ef 20 appreved by N *'C),

implem.ntati.n these.fmid eyele repair limits sh..l.

fellew the sam appreaeh as in TS 4.4.5.4.a.1O.a-,

4.4.5.4.a.l0.b, and 4.4.5.4.a.1g.e.

(2) z ~EISt value pr.vidd by the N. in GL 95 05 as siupplemented.

BZAVEIR VALLEBY UNIT-! 2 ~ 3ý4t--4 4 l4d Ame mnmnndment No.

z i0-1 0

UREI3LLN REQUIREnMRth tE (. ntinu.d)

b. The steama gcrnzrateLr shall be detefrmineed GPERABLE after emplet-In. the cirrcpnding . .r a.t..n. (plug rcpair all tubes cimcccding the plugging cr rzpair- limit) rcguiarrcd by Table 4.4-2.

a. within :15 days fzellewing the ee~mpletien ef eaeh enzri inspeetien ef steaff generater tu-bes, the ~rc uz plegged er- rzpairrzd in eaeh steam generater sihal-3 ibe submitted in a Speeial RepeLrt inac3ranee with 109 CFR

.0.rt

b. The mplet. results .. ef the stea g.n..rattr tube ad sleeve

,i.azrvice inspeetien shall be stubmitted in a Speeial flzper in aeeerdanee with 10 CTR 50.4 within 12 mznths fellewing the eempletien Cf the inspeetien. This Speeial Repert-shall inelude .

1. Ntffber and extent ef tubes and sleevz- npzcted.

2. Locatien and perccnt Cf wall thieimess penetratien feL eaeh inelieatien ef an imperfeeti--n.

3. idtntifi.ati.n Cf tubes plugged repaired. .r

e. Results Cf steam ge.rater n tube inspetien, which fall int Catgery G 3 shall be r.p.rted tC the G .. pursuant tC Speeifieatien 6.6 prier tC resufftien Cf plan eperatien. The written repert shall previde a deseriptie-n Cf investigatiens eendueted te determine the eatise Cf th tube degradatien and eerreetive measures taleen tC prevent reeturrenee.

d. Per implementatien Cf the veltage based repair eriteria tC tube plate

.upp.rt inter. . .tien.,netify thc Commission prier tC returning the steam generatecrs tC seravicC (?TGDE 4) shculd any Cf the fellewing eenditien

1. If estimated leakage based efl the prejeetd end Cf eyeic (Cr if net praetical, using the aetual measurz end Cf eyele) veltage distributien ezxczcds the leak limit (determined frefm the -in basis dese

.ico ealeulatien fer the pestulat Ad mVAin- ste-lin breakc) fer the next Cperating eyele.

fP.PZAt1TXT XI72'T.T.1~r TTNT'P3 : 3ý4L44 4l4e * ,-] .- TT 11 n 1

REACTOR COOLANT SYSTEM

2. if eiretimfercntial cracke lilee indieatiens are detected at the tubc suppert plate int"e.r. tien" .

3. If indicatins .areidentified that ex.tend by.nd the eenfines ef the tube.upp.rt plat-.

4. If indicatie.c ar- identified at the ...

pp.rt plate elevatiens that are attributable te primary water ote- crrcien cr-aeking.

5. if the ealeulated eenditienal burst prebability ba~c en the prejected end ef cyeic (er if net practical, using the aetual measureda ndl ef cyeic) veltage distributien emed 1: X ntify the Cezmboiss nQ~

and previde an assessment ef the safety signifieamee ef the eccurrencc-.

BEAVFER VALLEY UNIT 3ý4 14fAmendment Ne. 1:01-

INZ;pBECTBD D:Tj;CINBRVICEB I-NZPECTIO-N Prrcsrviee inspeetien -WeYe Ne. ef Steam Coneraters per Unit Thr~ee T-Three First Inserviee Inspocticn -M4 -we Seeend S ubsequent inserviee Inspeetiensa -G qe Table Neta-tio 1.Tho in.orico inspeetion may be limited te ene steam generater en a rotating seheduile encopaa 9; ef- the tubes if- the roaults of the first or previeus Inspeetiens indieate that all steam generaters are performing in a like manner. Net^ that tmder- aom eL circmtanees, the eperating enditiensa in ene er more steam genermaterms may be feund to be moere severe than these in ether steam gcneraters. Under sueh eiretfastanees the sample acquonee n. shall -A be modifiod to inapect4 ,&ýthe y -

moat

,J.- ý.

severe 9

conditiens.

-inoorvico i-nspetien shall be inspeeted. The third and nsaeer- ans a euld IF81 ew t- e ns"ruet, SR Eteserlile PTRWmYp UrzTrNUP TZMT-c 0 084847-

4-SABLE L4.

STEAMJ GENERA~TOR TUBE IUZPEGTIOGN A

i . m ....

s = ' whe~re n is the nth-abr ef steam caznerateLra insigeteel elriner -n ...-- etien.

-f T!)m, T7,T, Tr *T r -t 3L4 4 16 Amenme~mnt Ne. 101

REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATIO 3.4.5 SG tube intearity shall be maintained ANM All SG tubes satisfying the tube repair criteria shall be pluaaed or reoaired in accordance with the Steam Generator Procaram.

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

ACTION:

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

a. With one or more SG tubes satisfying the tube reoair 4 A ~ y 4 - v=A wq~r 4

=~r A =? r,. +-1, 4-1,~

Steam Generator Program:

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

2. Plua or repair the affected tube(s) in accordance with the Steam Generator Proaram prior to entering MODE 4 followina the next refueling outaae or SG tube

b. With Action a not being comoleted within the specified cormnletion 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

ý,I SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

qTRIVTT.T.ZTTl! 1n TTTRVMPTTrr Verify SG tube integrity in accordance with*.,n a f- sbtr'X 61 idýI 4 i t- sm .qht~-ckm (~r. ie-n rpt-gy Py-nrr 4.4.5.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugaed or renaired in accordance with the Steam Generator Proaram prior to entering MODE 4 following a SG tube inspection.

BEAVER VALLEY - UNIT 2 Amendment--No-,

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

a. No pressure boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

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

d. 10 gpm identified LEAKAGE.

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

ACTION:

bd. With any Reactor Coolant System operational LEAKAGE greater than any ene ef the a ....* nlimits, exnludingffor reasons other then pressure boundary LEAKAGE or primary to secondary LEAKAGE, reduce the LEAKAGE rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months er bc in at least 1OT CGTýMBY within thz next 6 heurs and in COLD .... T..W0 within the fellewing eh. With the reauired action and associated completion time of Action a not met. or with eay-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 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. 1-047 1

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)(3) least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during s .teady state eperaten.ft )

C- Verifying primary 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 .iz (2) Not required to be performed in MODE 3 zr 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months establishment of steady state operation.

(3) Not a)plicable to nrimary to secondary LEAKAGE.

BEAVER VALLEY - UNIT 2 3/4 4-20 Amendment No.-64 I

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 comoletion of an inspection performed in accordance with the Specification 6.19, Steam Generator (SG) Program. The report shall include:

'i3 a*. The scope of inspections nerformed on each SC in....l...

4 b. Active degradation ,as cf-nd inR, ssarized Water "Preii

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 or repaired during the inspection outaae for each active dearadation mechanism.

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

U. The results of condition monitoring, including the results of tube nulls and in-situ testing.

h. The effective nlugging percentaae 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* sA report shallbe submitted within 9e days after the initial entryo into MODE 4 following completion of an rinspection performed in accordance witht the Specification C6.19, Steam r~enerator Program,~ when voltage~ based alternate regpair criteria have been applied. The report shalli include information~ des~cribed in Section 6.b of Attachment 1 to Generic Letter 95-05, "'Voltage-Based ~Repair Criteria for Wftesting~house Steam ~Generator Tubes Affected by~ Outsid Qiameter Stress Corrosion Cracking ..

3, For imp~lementation of the voltagre-based renair criteria t t~ube- sunnTort plate intersections, notify the Commissionpro to returning the steam generators to service (MODE 4) should any of the following conditions arise:

,1. If estifft44ýed leakaeFe based en the 13ica4eet= ed enelýýeý evole (er- if not: praetjeal usineF -the a e t ual - end ef

,evele)- veltaae' distrib4ti efi ý e:;rpFaFaF4.0 r--he +F-laR +4! a::[:

-(dete ined frefft the-lieens inff baý;is daaeý-'e6leu atien" tlgp ovstial6ted fftain 'steamlin e bre6irg! fsr-th6 'next'66erat 1a.aIf circumferential crack-like indications are detected at the tube support plate intersections.

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

ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (continued)

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

kc.If indications are identified at the tube suDrort Dlate elevations that are attributable to orimarv water stress corrosion cracking.

"F S. i-tehus=ncbaiiybs I-the prajeet6d end efz eve-le fer if ne't praetieal. using t aetdal Ffteaiýýved -end 6f X An'sessme ef the, safetv eeeýiiarenee.

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 2A Amendment No. I

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

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

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

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

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

a. Provisions For Condition Monitoring Assessments Condition monitoring assessment means an evaluation of the "as found' condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubina 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 outaae durina which the SG tubes are inspected.

plugged, or repaired to confirm that the performance criteria are being met.

1. Provisions for Performance Criteria For SG Tube Integritv SQ tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakaae, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam aenerator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and desian basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady K

state full power pressure differential oneration and.

primary-to-secondary except as 'permitted thrcouah apelication of the alternate re9air criter discussed in Sn~ecification 6.19.c.4. a safety factor BEAVER VALLEY - UNIT 2 6-27 Amendment No. 1-2-G

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.

Apart from the above reguirements. additional loading conditions associated with the desian 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 sianificantly to burst or collapse. In the assessment of tube intearity, 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.

When alternate repaixr criteria discussed in

.. ecification 6.19.c.4 are applied to axially oriented outside diameter stress corrosion cracking at tube

,pupport on e or more locai ns - under plate indications the rbb-t ipostulated fbrtonmain st eam shal be less than -xlOa.

+line br eak conditions

2. Accident induced leakaae nerformance 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 leakage rate for an individual SG.

6Leakaae is also not to exceed ab-d1cm per SG

. ....ifi.d in' LtCO. .. 9.... except durincg a ý SG *!*tube 7i p,:Ure *r for specific tyves of degradation at specific locations as described in Tzchnigal Specification 6.19.c.4.

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

c. Provisions For SG Tube Repair Criteria

1. Tubes foQund by inservice inspection to contain a f law 151 *in a non-sleeved region with a depth ecaual to or

-.. exceeding 40% of the nominal tube wall thickness shall be plucraed or repaired -except w;hea 3ltzrnat6 tube yepairE eriteria tzzrmitted byif pernitted to remýtain in pervice throuah anplication of the alternate repair rzriteria discussed in -teehrical sSpec ifications

2. Sleeves found by inservice inspection to contain flaws with a depth ecrual to or exceeding the following percentaaes of the nominal tbe-sleeve wall thickness E: shall be plugged:

PTJ ABB Combustion Engineering TIG welded sleeves -%

Westinahouse laser welded sleeves 25%

3. Tubes with a flaw in a -sleeve to tube 0oint that 15 pccurs inj the sleeve or in~ the Qoriginal tube -wall of the joinit shall be plugcged.

4. The followihng alternate tube repair criteria may be applied as an alternativIe -to the 40% depth- based iriteria pof TechnicaltSpec~ification 6.19.*.1:

44-Tube Sunport Plate Voltaae-Based Repair Criteria Tube Supmort Plate Plugging Limit is used for the disposition of an alloy 600 steam generator tube for BEAVER VALLEY - UNIT 2

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) continued service that is exneriencina predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube sunport plates. At tube support plate intersections, 14.a: the nlugaina (repair) limit is ba=sd 'en m*aintainina a) Steam aenerator tubes, with degradation attributed to outside diameter stress corrosion crackina within the bounds of the tube support plate with bobbin voltaaes less than or eaual to 2.0 volts will be allowed to remain in service.

b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion crackina within the bounds of the tube support plate with a bobbin voltaae areater than 2.0 volts will be repaired or pluaged, excent as noted in 6.19.c.4.c below.

14. e c) Steam aenerator tubes, with Indi cations of potential dearadation attributed to outside diameter stress corrosion crackina within the bounds of the tube suport plate with a bobbin voltaae areater than 2.0 volts but less than or eaual to the upper voltaae 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.

14 .4[, d) Steam aenerator tubes. with indications of *otential degradation attributed to outside diameter stress corrosion crackina within the bounds of the tube suPoort plate with a bobbin voltage areater than the unner voltaae repair limit (calculated according to the methodoloay in Generic Letter 95-05 as supplemented) will be plugged or repaired.

BEAVER VALLEY - UNIT 2

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued)

e If an unscheduled mid-cycle inspection is 14.c performed, the following mid-cycle repair limits applv 4... instead of the limits specified in 6.19.c.4.a.

6.19.c.i4.b., a-ý6.19.c."4.c and 6,19,c4.cd.

The mid-cycle reoair limits are determined from the following ecuations:

VSL vMURL= SL 1.0+ NDE +Gr (CL-A)

SCL XCL - At VMLRL =VMUR - (V)n - CL

= uooer voltaae repair limit

= lower voltaae repair limit

= mid-cycle upper voltage repair limit based on time into cycle

= mid-cycle lower voltaae repair limit based on and time into cycle At = lenath of time since last scheduled inspection during which and were implemented CL = cycle lenath (the time between two scheduled steam aenerator inspections)

V... = structural limit voltaae r= averaae arowth rate per cycle lenath NDE = 95-percent cumulative probability allowance for nondestructive examination uncertainty (i.e.o a yalue of 20-percent has been approved by NRC) The NDE is the yalue provided by the NRC in GL 95-05 as supplemented.

14.c implementation of -these mid-cyCle repair limits shoQuld follow the same approach as in .Specification 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 oblective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present alona the lenath 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 5--3 Amendment No.

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued) outlet, and that may satisfy the aDrlicable tube repair criteria. The tube-to-tubesheet weld is not part of the

14. d tube. Prrevieus- dThfeztns er impr-cfeeticns in4 r 4.=--rd bm 23evnaar net egnsideged In tubes rapAir-ed

,by sleeving. the portion of the ogriinal tube wall between the sleeve's loints is not an area reauiring re-inspection.

In addition to meeting the reguirements 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 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 which inspection methods need to be employed and at what locations.

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

2. Inspect 100% of the tubes at sequential Deriods of 60 effective full power months. The first seguential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall onerate for more than 24 effective full power months or one intrerval between refueling outae's (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 -ower months or one lntervajl 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.

9F, ine gete re cz ggl-WAER VALLEY - UNIT 2

ADMINISTRATIVE CONTROLS STEAM GENERATOR PROGRAM (Continued)

4. I~dications left in service as a result of apolication of the tube sumport plate voltage-based repair 10 *riteria (6.19.c.4) shall be inspected by bobbin-moil g*

nrobe during all future refueling outages.

Implementation of the steam aenerator tube-to-tube support plate repair criteria reguires a 100-percent bobb*!!incoil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-lea tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-lea tube support plate intersections havina ODSCC indications shall be based on the nerformance of at least a 20-percent random sampling of tubes inspected over their full

e. Provisions for monitoring operational primary to secondary

f. Provisions For SG Tube Repair Methods Steam aenerator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removinX the tube from service. For the purposes of these Specifications. tube pluggina is not a repair. All acceptable tube repair methods are listed I. ABB Combustion Engineering TIG welded sleeves. CEN 2. etnon02 andulCEN-629-rwPlAddendumevs 1WCAP-fp=la E:

BEAVER VALLEY - UNIT 2 f _-u Amendment-Ho-

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

Provided for readability only

_ECHtNICA BASESon for 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 4-1g 3/4.4.4 PRESSURIZER ................................ B 3/4 4-2 3/4.4.5 3/4.4.6 STEAM GENERATORS (SG) Tube Integritv REACTOR COOLANT SYSTEM LEAKAGE .............

B 3/4 B 3/4 4-2 4-3 I

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 5-1a 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. 1-4149=aa 1

REACTOR COOLANT SYSTEM BASES Provided forInomtnOly I

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 of fsite power condition to maintain natural circulation at HOT STANDBY.

One OrPR2'.LLB~ steamu generater in a nen iselated reacter eeelant loop prevides sufflieint heat; removal capability to remeve decay heat-after a r.a.t. r .hutdewn. The u o fr tw OPERABLE sta gdniratirs, combined with .. ether r.. ir.m.nt. f the Limiting Conditions for O*pratien *nsures ad....at. day heat remova capabilitioc for RCZ temperatures greater than 3591F if enr steam gmn.rater baecmc .noperable due t" "id"ratien-4.

single failur"n-*

Belew 3591F, daceay heat -*- -- eelaby the RIIR system.

The Curveillanca Requirements for inspeetien ef the steami generator-tubes ensure that the structural integrity of this pertion of the RCZ will be maintained. The pregram for inserrviee Inspectien of ste-am generater tubes is based-on a medificatien of Regulatory Cuide l.83, Rf-viojon 1. Inservice inspection of steam g.n-rat -"tubing is essential in order to maintain survaillanee ef the conditiens ef the tu-bes in the event that there is evidenee of mechanieal damage or pr.gr...iv. degradation duo to design, manuf-cturin; ......

inoaU. eenditiens that lead te coroo. inseorvle Inspection ef steam generater tubing also provides a means ef charaeter-izingth nature and eause of ainy tube degradation se that co~rrootlve measures ean be taleen.

TPhe plant is expeeted to be eperated In a manior stuch thatth seeendary coolant will be maintained within these parameter limito found to roosult in negligible eorrooion of the steam ganerater tubac.

if the seeendary coolant chemistry is not maintained within thec parameter- limits, localized eor-rocon. ma-y likely result in stres&

i-- .cracking. The exctent of crackeing during plant BEAVER VALLEY - UNIT 1 B 3/4 4-2 Change No. I-Q-9"1029 I

REACTOR COOLANT SYSTEM REACTOR Cfor Inforation Only.

BASES eperatien weould be limited by the limaitatien ef steam generater tube lealeage between the Primary Ccclant System and the Seeendar-y Czzlant System (primary te seeendary LEAKAGSE w 150 gallens per day per stea generater). isaintaining- a prnimary te see .nda-, LEAAE; less than this limit: helps czcr adeefate mar-gin te withstand the leade impesed during nermal eperatien and by pestulated acdn3 operating plants have dem.n.trat.d that primary te see.ndary LEAK&CE

,f 150 gall. n, per day per steam .an ge.nratLr readily .... be dtete ,

anld -an Leakage inzrco f this-limit will require plant shutdewn tmsehzduled inspeetien, duIng which 'the icaleing tubes will b iccated and plugged.r Wastage type defeets are tmlikely with preper ehemistry ef zscccndary eeelant, sueh as previded by All Velatile Treatment (AT). IHcweverr even if a defeet ef similar type sheuld dezvelep in. se-r.ic, it will-be feund during seheduled snzvc team generater tube examinaticnz. Plugging will r ^-b - -'ed -f all t.ub.. with imperfeetiens emczzding the plugging limit. Steam generater tub inspectiens ef eperating plants have demenstrated the capability ta reliably detect a wastage type defect that has penetrated 20 percznt f the criginal tube wall thieknes..

Whenever the results ef any steam generater tubing inzric inspectien fall inte Categery C 3, these results will be reperted te the Co oioný pursuant tee Specificatien 6.6 prier t. rofuaptien cf plant .p.rati"n. Su-h eases will be e.n.ider.d by the Comma'io on

,a ease by ease basis and may result in a requirement fer analysih laber-atory--examinatiens, tests, additional eddy current inspectien, and revision ef the Tm-hnical Speeifi-ati-n- , if n-ee..a..y.

3/4.4.5 Steam Generator (SG) Tube Integritv BACKGROUND Steam aenerator tubes are small diameter, thin walled tubes that carry primarv 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 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 Dart 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 SQ. 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).

Providedfor Information Only.

SG tube intearitv means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable reaulatory reauirements.

Steam aenerator tubing is subject to a variety of degradation mechanisms. Depending uoon materials and design, steam aenerator tubes may experience tube dearadation related to corrosion phenomena.

such as wastage, pitting. intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These dearadation mechanisms can impair tube intearity if they are not manaaed effectively. The SG performance criteria are used to manage SG tube dearadation.

Specification 6.19. "Steam Generator (SG) Program." reauires that a mrogram be established 'and imnlemented to ensure that SG tube intearity is maintained. Pursuant to Specification 6.19. tube integrity is maintained when the SG oerformance 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 Proaram Guidelines".

APPLICABLE SAFETY ANALYSES The steam aenerator tube rupture (SGTR) accident is the limitina 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 eaual to the operational LEAKAGE rate limits in LCO 3.4.6.2.c. "RCS Operational LEAKAGE." plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes that following reactor trip the contaminated secondary fluid is released to the atmosphere via safety valves. Environmental releases before reactor trip are discharged to the main condenser.

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

they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere includes primary to secondary SG tube LEAKAGE eauivalent to the onerational leakaae limit of 150 and per SO. For _accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 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 regulatory 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 conse-uences 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)

LCOProvided for I The LCO requires that SG tube integrity be maintained. The LCO also reguires that all SG tubes that satisfy the repair criteria be pluaaed in accordance with the Steam Generator Program.

Durina 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 length of the tube. including the tube wall. between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered Dart 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 SG performance criteria: structural integrity.

accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a marain 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.a.. 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 aiven structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural intearitv performance criterion provides cruidance on assessina 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 nerformance 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 intearity reauires that the primary membrane stress intensity in a tube not exceed the yield strenath for all ASME Code.

Section III. Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in

Providedfor Information Only.

the design specification. This includes safety factors and aDplicable design basis loads based on ASME Code.Section III. Subsection NB and Draft Reaulatorv Guide 1.121, "Basis for Plugging Dearaded Steam Generator Tubes". Auaust 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 leakaae does not C.. .........

iit: in........... . .f exceed 150 cod mer SG.

accident induced leakaae rate includes any primary to secondary The LEAKAGE existing prior to the accident in addition to Drimary to secondary LEAKAGE induced during theaccident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during olant operation. The limit on Moerational 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 oer day. This limit is based on the assumption that a single crack leaking this amount would not pronaaate 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 aenerator tube integrity is challenaed when the nressure differential across the tubes is large. Larae differential pressures across SQ 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. orimarv to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

The ACTIONS are modified by a Note clarifyina that the actions may be entered independently for each SG tube. This is acceptable because the recuired actions provide appropriate compensatory actions for each affected SG tube. Comolying with the reauired actions may allow for continued ooeration, and subseguently affected SG tubes are aoverned by subseauent condition entry and application of associated reauired 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 olumaed in accordance with the Steam Generator Program as recuired by SR 4.4.5.1. An 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 Program. The SG repair criteria define limits on SG tube dearadation that allow for flaw arowth between insoections while still oroviding assurance that the SG nar nr ~ o r--rit-a-~ri 1il ormiif rma1 t-n him smt1- Tri evr90ir t-ca

Provide ormation Only.

determine if a SG tube that should have been plugaed has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated arowth of the dearadation nrior to the next SG tube insnection. 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 eyvaluation 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 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 plugaed prior to entering MODE 4 followina the next refueling outaae or SQ inspection. This completion time is acceptable since operation until the next inspection is supported by the operational assessment.

b. If the reauired actions and associated completion times of ACTION a are not met or if SG tube integrity is not being maintained, the reactor must be brouaht 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 The allowed completion times are reasonable, based on operating experience, to reach the desired plant conditions from full nower conditions in an orderly manner and without challenaina plant systems.

SURVEILLANCE REOUIREMENTS Durina shutdown periods the SGs are inspected as reguired 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 acceoted industry practices.

Durina SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found' condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance The Steam Generator Program in conjunction with the degradation assessment determines the scooe of the inspection and the methods

i Provided for Infomto used to determine whether the tubes contain flaws satisfyina the tube ny 1

repair criteria. Inspection sco'e (i.e.. which tubes or areas of I tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Proaram 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) techniue capabilities, and inspection locations.

The Steam Generator Proaram defines the Frecuency of SR 4.4.5.1. The Freouencv is determined by the onerational 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 Freauencv 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 reouirements concernina inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled insnections.

SR 4.45.2 During an SG inspection. any inspected tube that satisfies the Steam Generator Proaram repair criteria is removed from service by pluaaing. The tube repair criteria delineated in Soecification 6.19 are intended to ensure that tubes accented 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 Proaram, 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 Freouencv of "prior to enterina MODE 4 followina a SG inspection" ensures that SR 4.4.5.2 has been comoleted and all tubes meeting the repair criteria are pluaaed prior to subiectina the SG tubes to significant primary to secondary oressure differential.

BEAVER VALLEY - UNIT 1 B 3/4 4-2a Change No. 1--0124=

REACTOR COOLANT SYSTEM 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 1

REACTOR COOLANT SYSTEM P tde frRedbiiy 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 ProvidedforReadabilityOnly.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

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

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

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is BEAVER VALLEY - UNIT 1 B 3/4 4-3f Change No. 1-027

REACTOR COOLANT SYSTEM ProvidedforInformation 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 through Any One SG Operating zxaienee at; PWR plants has shewn that sudden inereases in leakc rate arae eften preeursers ta ilarger tui failtires. mlaintaining an eperating LEAKAGE limit ef 150 gpd per steam generateir will minimize the petenta fer a large LEARACE event at pewer. This3 eparating LEAYSACE limit isz mere raotrietive than the eperating LEAYJr.CE limit in standardized tehni-al . p..ifi-ati-ns. 'Phis prvides add"iti.nal margin te accemmdatz a tube f*law... *ih might-grew at a greater than expeeted rate er unempeetedly axten eutside the thieleness ef the tube suppert plate. Thias reduced LEA*AGE limit, in. .njt".ti. n with a leak rate m'nitern rz previdas additienal asouranee that thi preeuror-- - --CE; will be detactad and the plant shut dew inR a ti*m4ly mannar.The limit of 150 aallons per day ner 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 leakaae through any one SG shall be limited to 150 aallons per day." The limit is based on oDeratina experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the imDlementation of the Steam Generator Program is an effective measure for minimizing the frecencnv nf steam aenerator tubej rurtures.

BEAVER VALLEY - UNIT 1 B 3/4 4-3g Amndmnt-ChaneNo. 2 i--1-029

n - i -Iý. -= :-ý- I I ý:=.

REACTOR COOLANT SYSTEM Providedfor 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

_. Unidentified LEAKAGETor_. identified LEAKAGE, er prim.la*-y te

....ndary LFAYAGCE 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 LFAKPGE, identified LEAKcE, er primary te sccondary LEAXAGE .a.. t be r.du-ed t. within limite within 4 heurs, the reacter must be breught te !ewer pressure eenditiens te reduee the sever-Ity ef the LEAKcACE and its petential eensequenees. The reaeter must be breught te HODPE 3 within 6 heurs and mODE 5 within 36 heurs. This aetien reducee the LEAKAGE.

The allewed Ccmpletien Times are reasenable, based cn.

eperating ecxperienee, te rceaeh the required -lan

.eenditiens frem full pewer eenditiens in an erderl JazmL and witheut thal' nging plant; systems. In MDE5"h pressurte trerses ating en the RC.P. are mu. h lewer, and further deterieratien is much less likely.

BEAVER VALLEY - UNIT 1 B 3/4 4-3h Amendmcnt Chane No. 14&l2..

REACTOR COOLANT SYSTEM Providedfor Information Only.

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS (Continued) a_. If any pressure boundary LEAKAGE exists or primarv 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 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 I An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor the containment atmosphere wnrll, Rtui -h rtAn-n-;nmtznt- eiimnn ot 'T'h 1 7 hnirr mennn t-nvri nnr of the leaka1 e 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 re uired only on leakaae detection instrumentation reauired 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 leakaae detection instrumentation which is inoperable or not reauired to be onerable per LCO 3.4.6.1.

ER 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 te secndary LFAKAGE is also measured by perfermanee ef an flZ water inventery balanee In eenjunetien wi ef fluent menitering within the seeendary steam 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.

Therefere, this SR is net required te be perfermed in M9EDES 3 and-4 until 12 h.ur. *f steady state operation near operating pressure have

-ben .tabli*+h*.The SR is modified by two notes. Note 2 states that this SR is not reauired to be Derformed 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 Amenem entChance No. 4-_Q2. I

REACTOR COOLANT SYSTEM Providedfor Information Only.]

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

SURVEILLANCE REOUIREMENTS (SR) (Continued)

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

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

An early warning f pressure betundary L&M2CE or unidentified L&ERACE i~provi-3 by the systems that monitor the eentainment atmespher radoacivity and the eentaii:Hant sump level. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months moniterring the leakage detcction system Is stuffieient to previde an eariy efr -i 9f inereaed RCZ LEACE. These leakage t deteetien ystcms are speelfied in LCe 3.4.6.1, "Lcealeage Dcteetien instrtuxnntatien."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. Note (1) states that the 12 heour ureveillanec Is rcquirede nly en leakage detcctien in^trum.ntatio.

required by LCO 3.4.6.1. This Nete allows the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitoring to be suspended en leakage deteetien instrumcntatlen whieh -ia -inoprable or net required te be eperable per LCOe 3.4.6.1. Nete (2) states tha this SR i*-r-,-uirrd to be perfermed during steady state eperatien.

Note 3 states that this SR is not applicable to primarv 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 primarv to secondary LEAKAGE is less or e-ual to 150 aallons ner 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 apolies to LEAKAGE through any one SG.

If it is not practical to assign the LEAKAGE to an individual 9G. 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 nrimary to secondary

Provided forInomtnOly LEAKAGE determination, steady state is defined as stable RCS pressure. temperature. power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal iniection and return flows, The Surveillance Freauency 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 recoanizes the importance of early leakaae detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical arab samnlina in accordance with EPRI.

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

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

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

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

BEAVER VALLEY - UNIT 1 B 3/4 4-3j A &n . Lt'h&rn0 No. _Q2i I

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

Page B-Il 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

!1 Providedfor 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 I 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ................ B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY ............................ B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ................... B 3/4 4-6 3/4.4.11 REACTOR COOLANT SYSTEM RELIEF VALVES ......... B 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS ................................. B 3/4 5-1 3/4.5.2 AND 3/4.5.3 ECCS SUBSYSTEMS ...................... B 3/4 5-la 3/4.5.4 SEAL INJECTION FLOW .......................... B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT .......................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ......... B 3/4 6-10 3/4.6.3 CONTAINMENT ISOLATION VALVES ................. B 3/4 6-12 BEAVER VALLEY - UNIT 2 B-II Change No. 2-Q-2a= I

REACTOR COOLANT SYSTEM qProvidedfor 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.

'ý44. em.~Fzf ST GSF.1. ýA.vImý1 Gne 9PEflABn6B steam generater in a nen iselated reaeteLr eeelant leep I revides suff ieicnt heat remeval eapability te rzrnzve deeay heat

-F;rreaeter shutdewn. The zurmn fer tzvc OPERABLE stea generaters, eembined with ether requirements ef the iitn Ccnditiens fer Operatien ensures adequate BEAVER VALLEY - UNIT 2 B 3/4 4-2 Change No. 2-Q-2-&031 M I

Provided for Information Only.

REACTOR COOLANT SYSTEM

Proposedchanges to draftpagefrom q Unit 2 LAR 173 (EPU) deeay heat r-emeval eapabilities for RCS temperatures greater- thanl 3592F if ene steam generator beecomzo inoporable due to single failure eensiderations. Bolew 359 0F, docay--, heat Is romeoved by the lH system.~

The Survillianco Reoquirements for inspeetien of the steam generater-tubes ensure that the struetural integrity of this portion of the RCS will be maintained. The pr.gram fr in".r... "i ".....nptin f stea.n generater tubes is based en a medifcto of RegtlateL-y Guide 1.383, Rovision 1. Inoovc Inspection ef steam generator tu-bing is essential in ordr tomitisurveillanco of the condi~tions ofth tu-bes in the event that there is ev~idenee of meehanical damage eor pro.roie degradation due to design, manufaeturing orrors, or inservico conditiens that loadl to ---- r--- =. Inservico- inopotien eof steam gonerater tubing alse prevides a means of eharaeterizing th a~tuire and eauso of any tube degradation se that eorreetive measures can bo talkon.

The plant is expocted to be operatod iin a manner such thatth seeendary coolant will be maintained within these parameter limit fomnd to reoult in negligible eorrocion of the steam generator tubes.

If the so..ndary"colant chemistry is not maintained within th-parameter limi~ts, localizod oroinmay likoely result i~n stress.

--r---... .ra.king. The extent of eraecing during plant op-ratl"n would be limitcd by the limitation of steam g-nerator tube l4al.ag.

between the Primary Coolant S.ystem anld the SC...ndar Coolant System (primar- to seeondary TmEAKCB = 150 gallons per day per stea genorator). Aaial cracks having a primary to socondary LEARACE less than thi limi~t during operati-n will have.an d at margin of safety to withstand the leads imp.."d during .p.ratin..d n al b pestulatod accidents. Operating plants have domenstrated tha pria~-to socondai-y LFAKAE of 150 gallor.s per day per- sto gonerater can readily be doeteeted. LFM2CE in emeeoo of this limit will require plant ohkutdown and an unscheduled Inspection, during which the leaking tubes will be bocated and plugged or repaired by obooving. The teeniical: bases for sleoving are doocribod inth approved vender reporto listed in ZuL-villanee Reoquirement 4.4.5.4.a.9, as supplemented by Westinghouse letter PM;OC 02 304.

Wastage type dofocts are unlikely with the all 'volatile treatment-(AAF) of seeendary coolant. Hewevor, even if a dofect of- similar-type ohould develop in sorvico, it will be found duri~ng ocheduld in--ervicoe steam generator tube oxaminatiens. Plugging or repair will be required of all tubes with imperfoctions oxcooding the plugging or repair limit. Degraded steam gonorator tubes may be repaired by the installation of sleeves which span the degraded tube sectien. A steam genorator tu-be with a sleeve installed meets the structural BEAVER VALLEY - UNIT 2 B 3/4 4-3 Change No. 2--142-031 I

REACTOR COOLANT SYSTEM Provided for In BASES

/A A e

" TII'm"4 Uf"f 1*ITrn**"m, r ' T-e n. F ---- * -. A i- ef tubes which .. are ,,ents nmt degraded, th"r-f-r4, the szleeve iccznidered a part ef the tube. The survei~llanee requirrefntc identify these sleeving methcdclegicc apprvevd fzr use. If an 3ý4.4.-

I GEE4:PRis fetmd (Gen-- .... t in installed sleeve te have threugh wall p.n.trati.n greater than e al

.r ...t the .plugging limit, the tube must be pl.gg..

plugging limit fer the sleeve is derived frem R. G. 1.121 anialysis whieh utilizes a 20 perccnt allewanee fer eddy eurrent unccrtai~nty in-determining the depth ef tube wall penetratien and additienal degradaticn grewth. Stea gccater tube inspeetiens cf eperatIng plants have demenstratc ;Z_ capability te reliably deteet-degradatien that has penetrated 20 pcreent ef the eriginal tube wallI thielmessa.

The veltage basecd repai~r 34mits ef these survillianee requi~rement (SR) implement the guidanee in GL 95 05 and are applicable enly t Westingheuse designed steam generaters (S~s) w~ith eutside di~aflter stress eer-es-e .,crackeing (ODZCC) leeated at the tube te tube suppcr plate i .... ,-..... The guidan* in GE 95 05 will net be applied to the tube te flew distributien baffle plate interseetieno.* 4th Iveltage based repair limits are met applieable te ether feLMS cf SG tube degradatien ner are they applicable te ODSCCC that eccurs at ether lecaticns within the SG. Additicenally, the repair-cr-itcpria.

apply only tC indi"atien. where the degradatien meehanism i-deminantly axcial 9DSCC with ne NDS detectable cralcko extending eutside the thickncsc ef the suppert plate. Refer te GL 95 05 fer additienal descriptien ef the degradatien mer-phelegy.-

implcrncntatien ef these Zfls reelu-ires a der~ivatien ef the veltage strueturai. limit frrm thec berst verses veltage empirlcal eerrelatlen and then the sbubsequcnt derivatien ef the veltage repair limit Erm the struetural ourvcilla-ncc) . limfit (whieh is then imnplemented by thi:

The veltage struct-ural limit is the veltage frem the burst-pressure/bebbin veltage eerrelatien, at the 95 perccnt prcicic interval curve reeduccd te aeeczunt fer the !ewer 95/956 cn telcranee beund fer tubing material preperties at 6500 F (i.e., thc 95 perccnt LTL curve). The veltage structural limit must be adjusted dewnward te aeccunt fer petential degradatien grewth during ant eperating interval. and te aeeeunt fer NDE unccrtainty. The upper veltage repair liiV3lIiC determined frem the str~uctural veltage cy app+/-y~ng tnc r~+/-~;;~.n~ ~~iz~n:

BEAVER VALLEY - UNIT 2 B 3/4 4-3a A -e-"hnt nl No. 1O-1-_2=_Q*l

REACTOR COOLANT SYSTEM BASES I Providedfor Information Only.

Proposedchanges to draftpagefrom Unit 2 LAR 173 (EPU) 3I44. STA GEEAZER ~..-..

(Gentigir.

whei-e-J reprczsents the allewanca fer degr-adatien grewth betwez Cf ar~-in tameasuremeant ef the bebbin eeil veltage. rurthar-diseucsien ef the assumptiens neecsa~ccy te deterrmi~ne the velta-ge repair limit are discussed in G16 95 05.

Safety analyses were perfer-med purcsuant te Generie Letter 9.5 05ý that J.bb, S..*

eeuld epeura* . ,s .*

witheut ".......,

effsite "4 -

deses.. .. bL.....

exceeeding... ,.a simall ^4 .. . "'.*

fr~aetien 'ef 10C 50.6 gudelines (cansi~daring a eencuirrent iedine spike), 10 CFR 50.67 (,pre acc-ident iedine spilee), and witheut ccntrel reem desaz i 1,urrent FR 50.67. Uamaimum The value ef the M*B iinducad ieak rate and a sumarýy ef the analyses are pr-evlded in Seetiein 15.1.5 3f the UFrCR.

The mid eycle equatien in SR 4.4.5.4.a.!'.d sheuld only be used during uhnpla.nne. inspe.t.. n. In w. i.n e y . .. data.urr.nt is acq-ired fer indt at..n. at the tube suppert plates.

GR 4.4.5.5 implements several reperting raramnts rraaammnded by GL 95 05 fCr situatiens which the NRC wants te be netified prier te rcturning the SC* te servica. FPr the purp. . .s Cf this reporting ra -emnt, leakage and eenditi~enal burst pr-ebabili4t~y lanb-ea... t e based en the as f-und v.ltag- dictributin rather than t-.

a.

pr"e jtd Ayl end Cf (EGG) vCltage di.tributi.n (refer tC GL 95 05 feLr mere infe~rmatien) when it is net pr-actical tC eemplete these ealeulatienzz using the prejeeted EGG veltage distributiens prier t returning the Scc tC servica. Neta that If icaleage and eenditienal burot pr-ebability weira calculated tieing the meaauraed EGG veltaga distributian f Cr the purpeses ef addressing the GL sactien 6.a.1 n 6.a.3 raperting criteria, than the raeeults ef the prejeeted SGc veltage distributien sheuld be previded per the GL seetien 6.b (e+

eriteria.

Whenever the results ef any steam generatev tubing ic~c Inspeetien fall IntC Categery C 3, these racults will be repeLrted te 4-,- O%. 4-smS a a = ,,%=.%S=a ý V ,I¢,IL VJ--

A*

pa nt:~S e e .. .. *. ]L*L,.Z*.

. Sue=

- [.;.,Z*J,*oZ..

.,.U Z eases J,,*.;,; V..aa WJ*.- 11J.bJ we;;* .. U.4;J.;.;.*.L*

,SaA.... ,,, .....

,--*.4- . -n.

LLL.---.-.".-* .ty-

a. ..a - ~~.b6a.44~ ~ I.. %&.J .. & .. ~...saa. ~... %,.a %. a.aS&afr~~ a.~a.,

_.Z *4=k .. .. . .

-. - -se ei Otea cre,,-a speeifieatuiens, if ne. ss.

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

Steam generator tubes are an integral Dart of the reactor coolant

Providedfor Information Only.

pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as Dart of the RCPB, the SG tubes are u1nigue 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 reaulator reauirements.

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

Specification 6.19, "Steam Generator (SG) Program," recquires that a proaram be established and implemented to ensure that SG tube intearity 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. Meetina the SQ 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 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 eaual to the operational LEAKAGE rate limits in LCO 3.4.6.2.c. "RCS Operational LEAKAGE.' olus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes that following reactor trip the contaminated secondary fluid is released to the atmosphere via safety valves. Environmental releases before reactor trip are discharged throuah the main condenser.

cEtivit level of DOSF EQUIVALENTU I-13 is assumed to e ai t EUF-a to th

PPrrovviddeddfoor Information Only.

-U I

Gide 1.183 arid %Wf7thin GDC 9 values.

- xzi zfe--r A71A 1 mr c - eýc

ýr-(-! Hý--ntcý ýrr9 tr;=;nqi pnt.c,- t-h,ý-n ý) qrrPR Ac,-

ium, fli4- (Z(, t-lihenýc!

~~they "rp r9 i--hin (jiqrýIiArrTe t-n i-Tio icý -t-n nctD,-Vud Thrnzi y t- n ccnd r - Sr.~e t-if FA~rP - i nI - t .h

)ni4r=ýt ir)n-4~ 1 -: Tt - nf 1c 0 f 4'l nr q'~,:A T~nun- - An n S('-.tubp-q retainthej-T str 0~~aa h OI L)A Iv it~ of vTnt~rp 'hczr

. pn;:; I r r r j I- ý-r ia ý-n,:ý I y --ý- -- --- 4:ý-- nUrqiiant- t-n C,--nt-rir- Jpi-t-sýyý 5-0c) tD dete-ýe hda~rv~Lhe raayimummainýtea-mline break (MSLB) incii-ed without imA~v - -. ~ aE~Thha Ž~m1 !C~U i~ithn1i47A -)ff~i~n~~

f f !z: i i-*- Hn irimary to5f, r ý (-)-f 10 (T7R 5 0 -E7 as ccurI e -,r P rý e-n 11 ;:4 t- n Y-i i

-xceerli -n(-T i- b ý- I im i t- -, s ide1 183 aD( wj-tbDut coatrolr-oom , , t-3cr(,,-(jincr (,DC-1c) IPh P-

-n leakage adds '2.'1 cmmto the total leakaga ap p-";ý j -i CCLdoant-induced h~i~Th ~ ~f~1~KTh A~ Ihe ~ AnaW~ 1 ~ ~A t-h,- --tparý is. Therefore, i the MSLB analvýds.

Jnit- 2 MSLB ha~ A ith~analy ~atii~nh~e iii~Thd~

includes Dr ma~

mary to secondary qGtube U,-,(-hargetobeatmosphere

~~n1Wa1 ~t ~ anal A eM~a 14nitf~f of ThO..

150

~ii and *~r -SQ per ~

,E-ýKAGF.e=ivalent to the omýrational 1 kacTe limit

~th~ari cdd~n d~d~Ikage which~ results~ ina

£sIEud leaka f .4 cmm.

Steam generator tube intearity satisfies Criterion 2 of 10 CFR 50.36(c) C2)ii).

L-CO The LCO reguires that SG tube integrity be maintained. The LCO also reauires that all SG tubes that satisfy the reDair criteria be pluaaed or repaired in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Provided forInomtnOly I

Generator Proaram repair criteria is repaired or removed from service by Plugaing. If a tube was determined to satisfy the renair 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 Dart 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 SG performance criteria: structural integrity.

accident induced leakaae, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

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

Tube burst is defined as. "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g.. opening area increased in response to constant pressure) accompanied by ductile (plastic) tearina 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 loading conc Lition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance rri-rin--n e-n~oil ri gatig a lower striirtiira 1 li it- or 1 imi i-leT burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal 9 -F4~A 1 4-4 al I MMAradatCon A r Lhe claication Sar %AhA of A

axal*

1 A hem 62 E'a,..

l W lo &

n~ rrnrn

a. pi Fa,.an degradation, the classification of axial thermal loads as p~rimary or eerpnnrarv lnAdA will ht- PuAluat-g-d on a gagg-hv-t-AR hi.qA - rrh~~

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.

qpnt1inn TTT 5Aervice Level A (normal nnerat-ina t-nndii-inor-m ;;nd Service Level B (upset or abnormal conditions) transients included in the desian specification. This includes safety factors and applicable design basis loads based on ASME Code.Section III. Subsection NB and Thraqft- PrenilM~-nyrlr Clitirl 1 191 11R; Qi _fn-r PI iimerni nr TflenTr.=rltq _(tgr Generator Tubes". Auaust 1976.

EPU The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions durina plant 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 gallons per day, This limit is based on the assumntion 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 aenerator tube integrity is challen-ed when the pressure differential across the tubes is large. Larae differential oressures across SG tubes can only be experienced in MODE i. 2. 3. or 4.

RCS caiontns are far less challening in MODES 5 and 6 than durina MODES 1. 2. 3. and 4. In MODES 5 and 6. primary to secondary differential oressure is low, resultins 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 renired actions provide aipropriate comoensator_ actions for each affected SO tube. Complying with the reguired actions may allow for continued operation. and subseauently affected Sn tubes are aoverned by subseauent condition entry and aOplication of associated recuired actions.

a. ACTION a aAplies if it is discovered that one or more SO tubes examined in an inservice inspection satisfy the tube repair criteria but were not hluiaed or repaired in accordance with the Steam Generator Program as reuired by SR 4.4.5.1. An evaluation of SG tube integrity of the affected tubeis) must be made. Steam fenerator tube integrity is based on meeting the SG oerformance criteria

f, -, - ' -- 11 7 ý ', ý:m' k- " , ý , 7, -1 Providedfor Information Only.

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 refuelina outaae or SG tube inspection. The tube intearity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated -rowth of the degradation prior to the next SQ 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 outaae or SG inspection Provided the inspection interval continues to be supnorted 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 sunmorted by the operational assessment.

b. If the reauired actions and associated completion times of ACTION a are not met or if SG tube intearity is not being maintained, the reactor must be brouaht 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 The allowed completion times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging Plant systems.

SURVEILLANCE REOUIREMENTS During shutdown periods the SGs are inspected as reguired 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 Program ensures that the inspection is appropriate and consistent with accepted industry practices.

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

Providedfor Information Only.

The Steam Generator Program in conjunction with the degradation assessment determines the scope of the insLection 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 tubina within the SG are to be inspected) is a function of existing and potential dearadation locations. The Steam Generator Proaram and the degradation assessment also specify the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morpholoavg nondestructive examination (NDE) techn'iue capabilities, and inspection locations.

The Steam Generator Proaram defines the Freauency of SR 4.4.5.1. The FreauencY is determined by the operational assessment and other limits in EPRI, "Pressurized Water Reactor Steam Generato

V-.-m -211 rw 4-lV 2, e" ~a rThi Ct'-im m Dvrtrrv#- i"ag:60 ,rvini-on existina degradations and arowth rates to determine an inspectin 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 concerning inspection intervals to provide added assurance that the SG nerformance 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 growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s).

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

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

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 tubes to significant primarv to secondary pressure differential.

BEAVER VALLEY - UNIT 2 B 3/4 4-3b Change No. 2 G 642:ý= 1

REACTOR COOLANT SYSTEM Poie o edblt ny 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 R I 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 1

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

I Providedfor ReadabilityOnly.

Proposeddraftpagefrom Unit 2 LAR 173 (EPU)

APPLICABLE SAFETY ANALYSES (Continued)

Pr-imary te seeendarry LEAKAG'E; is a faeteLr in the dese releases eutside

.. ntaInment resulting frm a M.... a.i.d.nt. To a lesser extent, ether aeeidents er transients involve seeendary steam release te the atmespheLro, sueh as a steamf geneorater tube rupture (CT-R) .Th loakage --

taiates the seeondarry fluid.

The MSLE- __ ... r- limit~ing feLr site radiation releases. The pimar-y

=-4 ~~~ ~ nfM*" ~ ~ ~ _~ -j- ~ ý~ A~ ~

to seeendar-y 'LEUMCE -assumed In the safety analysijo for th.e IIZL]

=4 =-A. 4 _ .. ._1An a.Jb,4U aeei~dent is desei-ibed in UFSAR SeetIen 1.5.The r-adiological eonooquonees ef a MSLB eutsido ef eontainmont was reanalyod In, support of the tube suppe3rt plate veltago based roepair rtoi stated in SR 4.4.5.4.a.19. Per this analysis, the thyroid doac was ee in.id*nt iedine spike ease. RCZ leakage was based on prejet.ion r-ather than on teehnieal spocifieation leakeago limits. The analyaja indi"ated that eff-ito d. . ..

would remain within r.gulatory . rit-r"i wi~th the assumoed pr-imary to seeondar-y loakeago (deseiribed in T=Z Sootien 15.1.5) should steam genorater tuibes -fail dot h d.pres.urizati.n apeioiatfd with a LB. ..

A similar analysias was peLrfoL-md using a eontrol room thyjrorid doac otf 30 rem as the eritorion. The eontrol room wa - --timed to -bemanual-ly iselated and pressurized at 12-30 mainutes for- a period of enc hour-, at whi"h time filtered omergan-y intae- would be automati.allyastarted.

The control room would be purged with fresh air at T-8 hours following" ...

r.l.as. ati-n. The analysis indicated that eontrol roomf doaoas wouldl raai withi~n roegulatore3y eLritoriAa with the asseumod rm tocndary loakage (d...rib.d -in UFSAR Section 15.1.5) team generator tubes fail duoe to the depr------i--tion shul

.a.. oe with a "ZLTB.

-soc Primary-to-secondary LEAKAGE is a factor in the dose assessment of accidents or transients that involve secondary steam release to the atmosnhere, such as a main steam line break (MSLB), a locked rotor accident (LRA) , a Loss of AC Power (LACP2, a Control Rod Ejection Accident (CREA) and to a lesser extent, a Steam Generator Tube Rumture (SGTRI. The leakaae contaminates the secondary fluid. The limit on the nrimarv-to-secondarv leakaae ensures that the dose contribution at the site boundary from tube leakae.e following such

-accidents are limited to appropriate fractions of the 10 CFR 50.67 Thh limit of 25 Rem TEDE as allowable by Regulatory Giiide I1.R8 .

limit on the primary-to-secondary leakage also ensures that the dose contribution from tube leakaae in the control room is limited to the 10 CFR 50.67 limit of 5 Rem TEDE. Among all of the analyses that release arimaryactivity side to the environment via tube leakage.

the MSLB is of particular concern because the ruptured main steam line provides a pathwav to release the primary to secondary leakage directly to the environment without dilution in the secondary fluid.

F 71-,Iý-.ý,---- .-

ProvidedforReadability Only.

Proposeddraftpagefrom Unit 2 LAR 17 t EPII)

Due to adoption of the voltaae based steam aenerator tube revair criteria oer auidance provided by Generic Letter 95-05. the safety analysis for an event resulting in steam discharae to the atmosphere conservatively assumes a 450 ctd orimary-to-secondarv LEAKAGE (150 cnd per steam aenerator) for all accidents other that the MSLB. The dose conseauences associated with the MSLB addresses an accident-induced leakage, which, ner 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 conseauent hiah differential pressure across the- faulted steam generator The maximum allowed accident induced leakaae is 2.1 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 Amendment-Chariq gji o . -1ý2 --Q ýI-Q

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

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

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

c. Primary- to- Secondary LEAKAGE through Any One SG Operating experienee at PWR plants has shewn that sudden Ine-eases in leak rate are eften preedrsers te larger tube failurzc. maintaining an eperating L3EAK&GSE limit e-f 150 gpd per steam generater will minimize the petential c a large LEMUNCE event at pewera. This eperating LEAKAGE limit

.i mere restrictive than the epcrating LEAKAGE limit in standardized teehnieal speeifieatienzi. This prevides additienal margin te aeeenmmzdate a tube flaw whieh might-grc*w at a greatcr than expected rate er uncxcpzctedly .xtcn.

eutside the thiclcness ef the tube suppert plate. Th~i reelueed LEAKAGCE limit, i zjtinwith a lek ra~te M.nit.ri.. .,-.,jra, prv*...de additienal a....ran. that thic preeiurcrexZC will be deteeted and the plant shut dewn in a timely mann.r.The limit of 150 aallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Proaram Guidelines. The Steam Generator Proaram operational LEAKAGE performance criterion in NEI 97-06 states, '"The RCS operational primarv to secondary leakage through any one SG shall be limited to 150 aallons per day." The limit is based on operatinc" experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in coniunction with the implementation of the Steam Generator Progaram is an effective measure for minimizing the freauency of steam generator tube ruptures.

BEAVER VALLEY - UNIT 2 B 3/4 4-4g AmendmentCan= No. 19-02-031 I

REACTOR COOLANT SYSTEM BASES Pr ovidedfor Information Only.

i 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 Aenm~en-Chane No. 1-12-0-31 I

RASEACS C Providedfor Information Only.

REA*CTOR COOLANT SYSTEM [

BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS

b. Unidentified LEAKAGETor identified LEAKAGE, or primary to s...ndary LEAAGE Iin 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.

t,secondary LEAKAGE s annet be reducd tolimiton within within a,. . .*s.

.*.L4.-

4 .a* heuros,

. a. J ...*a*v the

,J reacter a vv**.J.a....a. %

muot

.&... ab. .*

be*..L brouight

. *.4 ,..J&.1.,*

JI to

. fr

!ewer

. vv .. ..... .&*

pressure eenditiens te reduco the severity of the PLT.CA4AE

-f and the dent,. =*,, ,.o* eenseefucncca.

itsen petential :n T'-- The roeacter ...must ren r=asonablo -E71 1 ..

be brouight to 11ODE 3 within 6 heturs and MODE 5 withint 36 heurs. This aetien reducca the LEA2chE.

The allewed Completien Times are reasenable, based eon.

eperating experienee, to reaeh the requIred pan eenditiens from full pewer conditions in an erderly manneor and witheut challenging plant systems. in 119DE 5, the pressure stresses acting en the RO-PB are mfuch lower, and further deterieration is much less likely.

ab. If any pressure boundary LEAKAGE exists or primarv 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 *-6--the followina 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

BEAVER VALLEY - UNIT 2 B 3/4 4-4i Am :mentCha No. O-22,-031 1

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

SURVEILLANCE REQUIREMENTS (SR)

SR 4.4.6.2-_a An early warnin) of pressure boundary LEAKAGE or unidentified LEAKAGE is 'rovidedby the systems that monitor the containment atmosohere radioactivity and the containment sumo 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 orovide 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 reauired only on leakaae detection instrumentation required by LCO 3.4.6.1. This Note allows the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months monitorina to be suspended on leakaae detection instrumentation which is inoperable or not recruired 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. s. ndary .. 6EAYA-GE is al.. mea.sured by perfermanee ef an RCZ water inventery balanee in eenjunetien wit efflucnt menitering within the seeendary steam and feedwater systefmo.

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

Therefore, týh~i SR is net required te be perfermed in 11ODEC 3 and-4 utitl 12 heurs ef steady state eperatizn near epefating pressere have~

been etablished.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 establishing steady state operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

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

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

An early warning ef pressure beundary LEAKAGE er tmidentified L&*%ý.C

-. Xlevidd by the systems that meniter the eentaiwnmznt atmespher radiazivity and the eentainment sum level. The 12 heur meniteringW ef the leakage Eeteetien s~ystem is suffieient te previde an early

,Pro Ivid ed fo r In f r a i n O fy warning ef inereased RCZ LEAKAGE. These leakcage deteetien systems-are speeified in LCO 3.4.6.1, "Leakage Deteetien instrumen 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. N.t. (1) states that the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillanee is required enly en leakage deteetien instrumentatin required by LGO 3.4.6.1. This Nete allews the 12 heur menitering te be suspended en leakage deteetien instrumentatien which is incpzrb er net required te be eperable per LGO 3.4.6.1. Nete (2) states tha thisi SR is required te be perfermed 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 or e-aual to 150 aallons per day throuah any one SG. Satisfyina the Primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance 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 aallons oer 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 assian the LEAKAGE to an individual SG. all the primary to secondary LEAKAGE should be conservatively assumed to be from one SO.

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 LEAKAGE determination, steady state is defined as stable RCS pressure, temperature. power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal iniection and return flows.

The Surveillance Freauency 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-Secondarv Leak Guidelines'.

BEANER VALLEY - UNIT 2 B 3/4 4-4j Amendme h No. 1-W2_a_0_31 I

'N Attachment C Beaver Valley Power Station, Unit Nos. 1 and 2 License Amendment Request Nos. 324 and 196 Miscellaneous Changes

Attachment C to L-06-088 Page 1 of 1 Miscellaneous Changes Change Description Differs From Reason for Change RAI Response T The response to RAI Item 1 indicated that LCO 3.4.5.b would Yes As stated in the response to the RAI, MODE be modified for both units to use MODE numbers in lieu of terminology and numbers are used the MODE names. This change was not incorporated, interchangeably. However, because the existing Technical Specification (TS) format does not use MODE numbers in the action statements, the MODE names have been retained. These TS will later be replaced in their entirety by the improved TS.

2 The response to RAI Item 10 indicated that TS 6.19.d.5 (now Yes The term "full length" would have been incorrect 6.19.d.4) would be modified for BVPS-2 to use the term "full because the term was intended to apply to the length" instead of "100 percent" with respect to inspection of number of tubes required to be inspected rather tubes when the voltage based alternate repair criteria are than the portion of each tube required to be applied. This change was not incorporated, inspected.

3 The response to RAI Item 6 indicated that TS 6.19.b.2 would No This change has been incorporated, but has also be modified for both units with respect to leakage limits been supplemented to clarify that the 1 gpm limit described in the accident induced leakage performance does not apply to a tube rupture event.

criteria.

4 Provisions for SG tube inspections described in TS 6.19.d.2 N/A Revised wording would preserve the intent of the and 6.19.d.3 for both units have been clarified with respect to intended requirements when literally interpreted.

quantities of tubes inspected and inspection intervals.

5 Several paragraph numbering changes have been made in Yes These changes were needed as a result of addition to corresponding references to those locations for additions and/or deletions of paragraphs resulting both units. Some of the renumbered paragraphs involve from the RAI or EPU, and to correct numbering paragraphs discussed in the RAI response. Those involving oversights not previously discovered.

the RAI are TS 6.19.c.1, 6.19.c.l.b, 6.19.c.l.c, 6.19.c.l.d, and 6.19.d.5. These have been renumbered to 6.19.c.4, 6.19.c.4.b, 6.19.c.4.c, 6.19.c.4.d, and 6.19.d.4, respectively.

6 Provisions for SG tube repair criteria described in TS 6.19.c.4 N/A The references to license amendments are not for BVPS-2 has been revised to remove a reference to the contained in the current TS and not needed to license amendment that originally approved the use of voltage completely define the requirements. Records of based alternate repair criteria. Provisions for SG tube repair license amendments are readily available to trace methods described in TS 6.19.f.1 and 6.19.f.2 have likewise the history of the repair method approvals.

been revised to remove such references.

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-06-088, Supplement to License Amendment Request Nos. 324 & 196 Steam Generator Tube Integrity

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L-06-088, Supplement to License Amendment Request Nos. 324 & 196 Steam Generator Tube Integrity

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