L-06-132, Supplement to License Amendment Request No. 183 - Submittal of Final Proposed Technical Specification Changes

Supplement to License Amendment Request No. 183 - Submittal of Final Proposed Technical Specification Changes

ML062490200
Person / Time
Site:	Beaver Valley
Issue date:	09/01/2006
From:	Mende R FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-06-132
Download: ML062490200 (26)

Text

FENOC Fir;tEnergy Nuclear OperatingCompany Richard G. Mende 724-682-7773 Director,Site Operations September 1, 2006 L-06-132 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Supplement to License Amendment Request No. 183 - Submittal of Final Proposed Technical Specification Changes By letter dated April 11, 2005, FirstEnergy Nuclear Operating Company (FENOC) submitted License Amendment Request (LAR) No. 183 - Revised Steam Generator Inspection Scope, for Beaver Valley Power Station Unit No. 2 (Letter L-05-061, Reference 1). Revised markups to the proposed Technical Specifications and Bases were provided on January 27, 2006 (Letter L-06-013, Reference 2).

On July 19, 2006, the NRC issued Amendment 156 to the BVPS Unit No. 2 Operating License, authorizing the implementation of an extended power uprate. On August 3, 2006, FENOC provided a Supplement to License Amendment Request Nos. 324 and 196, Steam Generator Tube Integrity, for Beaver Valley Unit Nos. 1 and 2 (Letter L-06-119, Reference 3). Attachment A provides final proposed changes to the BVPS Unit 2 Technical Specifications, reflecting both LAR No. 183 and the changes resulting from the above two licensing activities. Attachment B, which proposes final changes to the Technical Specification Bases, is provided for information only.

The proposed final changes to the Technical Specifications do not affect the conclusions of either the supporting safety analysis or the no significant hazard evaluation provided in Reference 1. No new regulatory commitments are contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Gregory A.

Dunn, Manager, FENOC Fleet Licensing, at (330) 315-7243.

9o961

Beaver Valley Power Station, Unit No. 2 Supplement to LAR 183 L-06-132 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on September 1 , 2006.

Sincerely, Richard G. Mende

Beaver Valley Power Station, Unit No. 2 Supplement to LAR 183 L-06-132 Page 3

Reference:

1. Beaver Valley Unit No. 2 License Amendment Request No. 183 - Revised Steam Generator Inspection Scope, Letter L-05-061 dated April 11, 2005

2. Beaver Valley Unit No. 2 Supplement to License Amendment Request No. 183 Revised Steam Generator Inspection Scope (TAC No. MC6768), Letter L-06-013 dated January 27, 2006

3. Beaver Valley Unit Nos. 1 and 2, Supplement to License Amendment Request Nos.

324 and 196 - Steam Generator Tube Integrity (TAC Nos. MC8861 and MCS862),

Letter L-06-119 dated August 3, 2006 Attachments:

A. Proposed Technical Specification Changes - LAR No. 183 B. Proposed Technical Specification Bases Changes - LAR No. 183 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 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Changes License Amendment Request No. 183 The following is a list of the affected pages. Since changes related to LAR No. 196 (Steam Generator Tube Integrity) are subject to license amendment issuance, the most recent revisions proposed for that LAR (Letter L-06-119, August 3, 2006) have been incorporated in the attached pages. Additional markups related to LAR No. 183 are shown in strike-through/double-underline format.

Page 6-22*

6-22a 6-27*

6-28 6-29*

6-30 6-31 6-32

• This page is not changed and is provided for readability only

Providedfor Readability Only.

ADMINISTRATIVE CONTROLS Proposeddraftpagefrojm Uhn* 2 LAR 1 196 (TSTF-449 - SG Tube lntiegri09, PRESSURE AND TEMPERATURE LIMITS REPORT (continued)

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

6.9.7 STEAM GENERATOR TUBE INSPECTION REPORT

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

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

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

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

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

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

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

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

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

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

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

BEAVER VALLEY - UNIT 2 6 -22 Amendment No.

SMarkups Unit 2toLAR proposed draftpage 196 (TSTF-449 - SG COTROLSfromt ADMINSTRATVE ADMINITRATIV COTRLSt be Integ-rty)

STEAM GENERATOR TUBE INSPECTION REPORT (continued)

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

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

A-_R-p-o-rt-the follow _i'*_iiforma-ion -to the-N-RC- ws2i 9 ~

after achievinq Mode 4 following an outaqe in which the F*

methodoloqy was appliedk

a. Total number of indications. location of each indication, orintion of -e-a*!___ lcation, severi of each indication, and whether the indications nitiated from the sideroro-utsid__face

h. The cumulative number of indications detected in the Vubesheet r-eqon as a-fjunction of el eyvaeLi-on-within the tiub e sihvet c, The projected end-of-cycle accid -nducedi kage from t'besheet indic~a~timos_

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

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

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

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

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

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

II Providedfor Readability Only.

ADMINISTRATIVE CONTROLS Proposeddraft pagefrom Unit 2 LAR

=1 196 (TSTF-449 - SG Tube Inte'rity)

TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM (Continued)

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

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

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

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

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

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

b. Provisions for Performance Criteria for SG Tube IntegritV SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 6.19.c.4, a safety factor BEAVER VALLEY - UNIT 2 6-27 Amendment No.

ADMIISTRTIVECONTOLSfront Ult 2LAR71976(TSTF-449 - SG ADMIISTATIV COTROL I Markups toTube Integ~rity)draft page proposed STEAM GENERATOR PROGRAM (Continued) of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

When alternate repair criteria discussed in Specification 6.19.c.4 are applied to axially oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lxlO0.

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

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

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

c. Provisions for SG Tube Repair Criteria

1. Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in Specifications 6.19.c.4__or 5 5

2. Tubes with sleeves found by inservice inspection to contain *_ flaws in a sleeve (that are net in excluding the sleeve to tube joint-I with a depth equal to or exceeding the following percentages of the nominal sleeve wall thickness, shall be plugged:

ABB Combustion Engineering TIG welded sleeves 27%

Westinghouse laser welded sleeves 25%

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

4. T-p-ba _s upp-o r_* _pla* _-eY-0-1tta eca

-

• _re p-air_ _cr-jt-eria Thee f-ellewing alternate tube repair eriteria-may be applied as an alternative to the 40% depth based criteria of Technical Specification 6.19.c.I-Tube Support Plate Plugging Limit is used for the disposition of an Alloy 600 steam generator tube for BEAVER VALLEY - UNIT 2 6-28 Amendment No.

II ADMINISTRATIVE CONTROLS Proposeddraft SProvided forpage front Unit 2Only.

Readability LAR (Continued)

STEAM GENERATOR PROGRAM continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is described below:

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

b) Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts will be repaired or plugged, except as noted in 6.19.c.4.c below.

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

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

BEAVER VALLEY - UNIT 2 6-29 Amendment No.

ADMNISRATVE ONTOLSfroim Unit 2 LAR 196 (TSTF-449 - SG ADMINISTRATIV CONROS arkups to proposed draft page STEAM GENERATOR PROGRAM (Continued) I Tibe Integrity) e) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits specified in 6.19.c.4.a, 6.19.c.4.b, 6.19.c.4.c and 6.19.c.4.d.

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

v SL MURL -CLAt 1.0 +NDE +Gr. CL "

CL CL- A)

VMLRL =VMURL -(VURL- VLRL)(

where:

VURL = upper voltage repair limit VLRL = lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMURL and time into cycle At length of time since last scheduled inspection during which VURL and VLRL were implemented CL = cycle length (the time between two scheduled steam generator inspections)

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

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

5. The F* methodoloyas described below, m be applied to the expa d prtion of the tube in the hot-leq tubesheet region as an alternative to the 40% depth based criteria of Technical Specification 6.19.c.l:

) Tubes with no portion of a lower sleeve joint in the holt tesbesheet re-gimshall berpire rpiw qqed 1pon detection of any flaw identjfie within 3*.*0 inches -below theitp of the tu be!hea*_ within 2.2 inches below the bottom of roll transition, whichever elevation is lower, Flaws located below thi el__ie2 __rdmaininsexvicregqies__

size.

kTTes which haveany__pprtion_ _ leeJ*veint in the hot-leg tubesheet re ion shall be p luqqgA upon detection of avry flaw i dMtiedL-wi-thin 3.0 inches below the lower end of the lower sleeve joint, Flaws located qter than 3.0 inches- eow the lw eno the lower sleeve joint may remain in servicQ egaredless of size-

d. Provisions for SG Tube Inspections Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube BEAVER VALLEY - UNIT 2 6-30 Amendment No.

ADMIISTRTIVECONTOLSfroin IMarkups Unit 2toLAR proposed draft page 196 (TSTF-449 - SG COTRLSt ADMINISTRATIV be Atlegrity)

STEAM GENERATOR PROGRAM (Continued) outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection. In addition to meeting the requirements of d.1, d.2, d.3, an--d.4. _an d.L below, the I inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be 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 periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one interval between refueling outages (whichever is less) without being inspected.

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

BEAVER VALLEY - UNIT 2 6-31 Amendment No.

Markups to proposed draftpage from Unit 2 LAR 196 (TSTF-449 - SG ADMINISTRATIVE CONTROLS Titbe httegriý)

STEAM GENERATOR PROGRAM (Continued)

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

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

5. when th2_F_*meodo_1oqyhabaen-i-p-lee!* *_i/ispe¢ 100% of the inservice tubes in the hot-lee tesheet r-e-Wion-wjth the obective of dretgctnflawstha

_atisfy he pplicable tube repair criteria of Technica Sp~e*ifc~tion 6.19.c.5 every___24 effectije full power months or one interval between refueling outages (whichever is laessJ.

e. Provisions for monitoring operational primary to secondary LEAKAGE

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

1. ABB Combustion Engineering TIG welded sleeves, CEN-629-P, Revision 02 and CEN-629-P Addendum 1.

2. Westinghouse laser welded sleeves, WCAP-13483, Revision 2.

BEAVER VALLEY - UNIT 2 6-32 Amendment No.

Attachment B Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Bases Changes License Amendment Request (LAR) No. 183 The following is a list of the affected pages. Since changes related to LAR No. 196 (Steam Generator Tube Integrity) are subject to license amendment issuance, the most recent revisions proposed for that LAR (Letter L-06-119, August 3, 2006) are shown in strike-through/double underline format. One additional markup related to LAR No. 183 is annotated as such.

Page B 3/4 4-2*

B 3/4 4-3*

B 3/4 4-3a*

B 3/4 4-3b

• Provided for readability only

REACTOR COOLANT SYSTEM Providedfor 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.

One OPERABLE steam generater in a nen iselated reaeter eeelant leep prevides suffieient: heat remeval eapability te remeve deeay heat aftor a r** ctcr shutdewn. The requirement fer two OGPERLE team generaters, eefftbincd with ether requireomonts ef the iitn Conditiens fer operatien ensures adequate BEAVER VALLEY - UNIT 2 B 3/4 4-2 change No. 2-42-5031 1

Providedfor Information Only.

REACTOR COOLANT SYSTEM Proposedchanges to draft pagefrom IUnit 2 LAR 173 (EPU) docay heat romoval capabilitios for IZCC temperatures greater than a59 0 F if ono steam generater bocomos Inoporable due to single failure

..nsid.rati.ns. olow 3359 0F, docay heat +/-S removod by the nRn The Surveillanco Requirements for inspeetien of the steam genorater tubes ensure that the structural integrity of this portion of the RCZ will be maintained. The program for inservaico inspectien of stoam generator tubes is based -an a modifieation of Rogulatory Guide 1.83, Fovision i. inservico inspeetien of steam generater tubing is essential in ordo--r to maintain survoillanco of the conditions of the tubes in the event that there is ovidonco of mochanical damage or progress-iv degradati, n due to design, manufaturing errors, or inservice conditions that lead to.orrosion. ins.rvi..

• n f insp"" ti.

steam gonorator tubing also provides a means of characteri zing the nature and cause of any tube dogradation so that oreroative measures can be taken.

The plant is . xpo.t. d to be operatod in a manner such that the socondary coolant will be maintai~nod wi~thin these parameter- limit found to result in negligible corrosion of the steam gencratsr tubes.

if the s pcondlant coln ch.mistry is not maintained within these parameter li~mits, localizod corrosioen may likely result in stress corros--io crackeing. The extent of cracking during plant eporatien would be limitod by the limitation of steam gonorator tube leakage between the Primary Coolant System and the Socondary Coolant Systo (primary-to socondary LEAKAGE -150 gallons per day per steam genorator) .Axial crackes having a primary to socondary LEAKA RE less-than this li~mit during operatien will have an adequate margin of safety to withstand the loads imposod during normal eperatien and by, postulatod accidonts. eporating plants have domoinstratld ta primary to socondary LEAKAGEl of 1SO gallons per day per steam gonorator can readily be dotoctod. LEAKACE in oemcoss of this limit-will require plant shutdown and an unschodulod inspoction, during which the leaking tubes will be locatod and plugged or repaired b slooving. !The tochnicalI bases for slooving are doscribod in the approvod vender roports listed in Survoillanco Requirement 4.4.5.4.a.9, as supplemented by Westingheuse letter PENOC 92 394.

Wastage type dofocts are unlikoely with the all volatilo treatment-(AVT-) of secondary coolant. llowever, even if a dofoct of similar type should dovobop in sor-vico, it will be found during schodulod insor~vico steam gonorator tube examinations. Plugging or repair will be required of all tubes with imporfoctions oxcooding the plugging orýj-reopair limit. Degraded steam gonorator tubes may boroeppairod by t1ho installatien of sleeves which span the degraded tube soction.A steam gonorator tuibo with a sleeve installed meets the structural BEAVER VALLEY - UNIT 2 B 3/4 4-3 Change No. 2--Gi2-_z31 I

REACTOR COOLANT SYSTEM Providedfor Information Only.

BASES

/4.4.5 ST-1AM CENT.RATORS (Conti~nued

... uir.m.nts

.f tubes whieh are net dcgradcd, thrrftcr, th....

dol is.. n.. hred a part the .f tube. The survcillan r i identify thcse sleeving mcthedelegies apprevad fpr use. if an installed sleeve is~ feund te have threugh wall pcnctratien greater than cr equal te the plugging limait, the tube must be plugged. The plugging limnit fer the sleeve is derived frem R. C. 1.121 analysis whieh uitilizes a 20 pcreent allewanee fer eddy current uneertainty in determining the depth of tube wall pcnetratien and additienal dcgradatien grewth. Steam generatcr tube inspeetiens ef epcrating plants have demenstrated the capability te reliably detcct degradatien that has penetrated 20 perccnt ef the criginal tube wall thiekness-.

The veltage based repair limits ef these ziurveillanee reguirements (SR) implement the gui~danee in GL 95 05 and are applieable enly te, Wczjtingheuse designed steam generaters (S~s) with eutside diameter stress eerresien cracking (GDSCC) leeated at the tube te tube suppert plate intcrsctienso. The guidancc in CL 95 05 will net be applied t-the tubc te flew di-tributi-n baffle plate intersecti.n.. The

...ltag based repair li..it .

are net appli.abl. t. ether ferm. ef SC tu-be degradatien ncr are they applicable te ODCCCG that eeeurs at ether lc.aticn. within the SC. Additi.nally, thc rcpair "ritcria apply only tc indieatiens where the degradation meehanisfm i-a dominantly axial ODCCC with no N4DE d.t.table er-ak. extending outziide the thickenccz of the support plate. Reafer to CL 95 05 fo-r additienal dczicriptien cf the dcgradaticn morphology.

implementatien of these CRs requires a drivati.n ,f the vltagc structural limit frm the burst vcrstus voltagc ... .

pirical . rr.lati. n and then the subsequent dcrivatien cf the veltagc repair limit Erem the structural limit (which is then implemented by this survcillan*e).

The v.ltag. structural limit is the vtltag . from the burst prcssurc/bebbin vcltage corrclatien, at the 95 pcrccnt predicticn interval curvc reduced to account fcr the lcwcr 95/95 p.r..nt telcrancc bouind for tubing material prepcrties at 65()'F (i.e., the 95 pcrccnt LTL curvc) . The voltagc structural limnit must be adjusted downward to account for p*otntial d.gradati.n growth dur-ing a cpcrating interval and tc acccunt; for NDE uincartainty. T-hae u ppcr Nlcltage repair limit,;~L is determined from the structural otg li-*mi - b-y applying thclqfclowngcatieo.

BEAVER VALLEY - UNIT 2 B 3/4 4-3a Amendmentjhamg No. 1Q-2-_0_31

'I REACTOR COOLANT SYSTEM BASES I Providedfor Information Only.

Proposedchanges to draftpagefroui Unit 2 LAR 173 (EPU) 9/4.. STEA ,(G-entinu,.*.* ',*"

GBBRAGR .. -

where-V represents the allowanco for dogradatien growth between of errer in- tho masuromnt of the bobbin coil voltago. Further discussion of the assumptions nocossary to detor-mine the veltage

-- ~~~ ~ ~ ~ ~ - ~ ~ . r-I-I- --- 3!1- - ' T n Safety analyses were performed pur-suant to Conerie Letter 95 05 t deter-mine the maximumn MSLB indueed primary to socondary leak rate that eould eeeur- without effsite deses emeding a small fraetion of 10 CER 59.67 guidelines (eonsidor-ing a eencurrent iedino spikoe), 10 CER 4-S9.6q -- ,* (pre r,, aocident

. ,"A -L ... Y - iedine

.fl.14~r spike), and without control . room 44-flfl.?," 4-4 deses

• 4 oxcooding 1:0 CFR 50.67. The eurrent value of the maximumn ?MLB inducod loake rate and a summ ary of the analyses are provided in I ......

The mid cyele oguation in SR~ 4.4.5. 4.a.4iu.ei Ofeulek only ne usod-during unplanned inspoctions in whieh eddy current data is ac-,uIrod"---

for indications at the tube support plates.

SR 4.4.5.5 implefmonts several ropor-ting roguromnts rocommofnndod by CL3 95 05 for situations which the NRC wants to be notifiod prior to returning the S~s to sorvico. For the purposos of this reporting requirement, loaleago and conditional burst probability can be r- , . 1 -- -

V . - "

projoctod end of cyclo (BeeC) voltago distribution (refer to CL 95 0-5 for mere infermation) when it is not practical to comploto these calcuilatioens using the projoctod EGG voltago distributions prior to reoturning the G~s to sorvieo. Noto that if loakeago and conditional-burst probability were calculatod using the moeasurod EGG voltago distribution for the purposos of addressing the GL soction 6.a.1 and 6.a.9 reperting eriteria, then the results of the prejected EGG

-voltago distribution should be providod per the GL sootion 6.b (e) eritoria.

Whenever the results of any stoamn gonorator tubing Rn--rvi--

inspectien fall into Catogory C 3, these results will be roportod to

-2 -

the Commission pursuant to Speeificatien 6.6 prior to resumption of-plant eperatien. Such eases will be eonsidorod by the omsino a ease by ease basis and may result in a r--uIromoe-nt for analysis, laboratory oxaminations, tests, additional eddy curront inspectien, t..&+/-JA...& .5..'.. V.3. S..~.V&&AA ~ *~ ~.'.~bJJt..4.. 1 3/4.4.5 Steam Generator (S-G)Tuhernte-irt BACKGRQOND St-ea_*qgenerator tubee__ -s*!a__*£_aneter, thin w~a-11-__tubes tha arry primary coolant throuQh the primary to secondary heat exr~h~naer~ - The SG tubes have a number of imnortanti sfetyv fiinctions, Steam qenera tor tubes are an-inteqiral _pLart-of thhereactor

I Providedfor Information Only.

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

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

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

Steam generator tubing is subject to a variety of degradation mechanisms. De iq__upon materials and des-in, steam _enerator tubes may experience b egrdation related to corrosion phenomena, such as wastage, pitting. intergranular attack, and stress corrosion cracking. along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The S. permance criteria are used to manage SG tube degradat-ion.

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

There are three SG performance criteria: structural integrity.

accident indcd e. and operational LEAKAGE. The SG performance criteria are described in Specification 6.19. Meeting the SG performance criterxia provides rea ble assurance of maintaining tube integrity at normal and accident conditions.

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

APPLICABLE SAFETY ANALYSES The steam generator tube rupture (SGTR) accident is the limitinq diesign basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes aounding primary to secondary SG tube LEAKAGE rate eaual to the operational LEAKAGE rate limits in LCO 3.4.6.2.c. "RCS Operational LEAKAGE." Dlus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes that followinq reactor trip the contaminated secondary fluid is released to the atmosphere via safety valves. Environmental releases before reactor rp a ds ed throuqh the main condenser.

For accidents that do not involve fuel dam ge. the primary coolant activity level of DOSE EQUIVALENT 1-131 is med to LCO 3.4.8, "RCS Specific Activity." limits. Pre-accident and

-concurrent iodine spikes are assumed in accordance with maplicable regulatory guidance. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these

I,- - . - A Providedfor Information Only.

i- m-n t-czn-. wi t~hin i-h i 1m it- c:rif 10 (rFR 9;0 7 q c:tiinnlim1t-nt-,zei -n, Req-ulaltory Guide 1.183_and within GDC-19 values*

heanaysis for des~igabsis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity(i.e.

they are assumed not to rupture) dthe stee to the Atmosphere is assumed to includ-eprimary to secndarvy SGtxube LEAKAGE exuivalent to the ooerational leakace limit of 150 a ner Sf.

pd However, an increased leakage aassumption is A_ lied in the Unit 2 RlC:Tl

= -n = ' x i' *. In suoDort of voltaQe based reDair criteria In sur)T)ort of voJ-taqe-JDas-ed--r-e.P-air criteria Dursuant 'Dursuant to Generic Letter 95-05. analyses were performed to determine the maxm] brn am 1 (UB r ondr earate thatc*o cgr withoutIIosite- excedir the limits of 10 CFR 50.67 as_ supp lemented9by Reg*lato lide 1 cntrolo o*_dQses exceedinGC-19

• An-a-ddit-ionaJ12_, _q

_sasssumed in the Unit 2 MSLB anal*yis resu1 ing from accident c~onditi ons. Therefore. in the MSLB analysis, the steam discharae to the atmosphere includes primary to secondary SG tube LEAKAGE

_e-q-_iva lent to the operational le ak eqe_limit of 15-0 qpd per SG-and an additional 21 gqpm which resultsj*n a total asssumed cident induc 1 eakaqe of 24 qpm Steamm generator tube in _ri_ satisfies Criterion 2 of 10 CFR 0.3-6 (c) (2-1il LCO F* INSERT The LCO regpirestha i so r~~uies thataIll The combined projected leak rate from all p~lu4gqed or reDpairedL alternate repair criteria (i.e., voltage based repair criteria and application of F*) must be less than the maximum allowable During an SG inspect fb-e steamline break leak rate limit in any one Gen*-rator Pr~ogainre byp I!ug ing.If a__ steam generator in order to maintain doses within the limits of 10 CFR 50.67 as

_buwa s not D_!_uqqgýe supplemented by Regulatory Guide 1.183 and within GDC-19 values during a postulated steam line break event.

In the context of t

• ntireiength of th.

maeto t. between weld is not considered part of the tube, ASG Ltube has tube integrity when it saifies the SG performance 6,rit9era. The SG ance Crieriaare ed in Spion 6.19. "Steam Generator Program," and describe acceptable SG tube pformance. The Steam Generator Program also provides the

• valuation process for determining conformance with the SG pe rftormance crieria*

There are three SG performance criteria; structural integrity, anyone of these criteria is considered failure to meet the LCO.

The staructural integritype rformance critrAo__rovi~es_ mar-lnof anfaccde under normal safety against tube burst or collapse

• • r of the SG tub all anticinated transients included in the desion snecification ll antic nated transients included in the-deýiqn speci i a

I Providedfor Information Only.

Tube burstLs__efined as, "The gross structural failure of the-tube wall. The condition tp 1ig1y corresp s to an ulnst* _le__penin diplacement (e. ,, opening area increased in response to constant pressure) accomp anied by ductile (plstiL ) t earing of the tube matexri-a1__at the ends of the _deqrad-a-tion." Tuhe__cgl-a s~eisde~fi s, "Foricthe load displacgement curve for a qiven s ructure, collapse

_ccis__*a a the top of the la~d __di acement curve where the slope of the curve becomes zero." The structural inteqrity performan c ion prpxjdes_ qui-gnce on asesing loa thathave asgqnificant effect on burst or collapse. In tiiat context, the term "siqnificant" isdefined as "An accident loading n ondition other than differential pressure is c-ai-dared sigqi-fi-aa-nt when the addi~ti such loads in the assessment of the structural inteqrity performance criterion coul c!e ailower structural limit or limitin' burst/collapse condition to be establishea " For tube integrity vha tions. excet felde-qraati _a i thermal loads are classified as secondary loads. For circumferential dgradxation. the classification of axialdterm lads asprimary or secondary loads will be evaluatd on a case-bL-case basis The division between primar and secndary c!assifications will be based Qdetaiaayssn lor t~esting

_tructurajl int egrity-_re-qugires that the _rymembrene stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normalopgexatin conditions) and Service Level B (upset or abnormal conditions) transients inclu-ded in the design specification, This includes safety factors and ap-p Ii cable de s-d hae'i--loaids-_ased on ASME CoedaSeion L I III Subsection NB and Draf t; Reulp Guide 1,121, "Basis for Plugging

_Daxde-d-Steam Generator Tubes," u'ust1976 The accident indAa-pIIkda-qeperformance cri nn ensures that the primary to secondar LEAKAGE caused bva desi gnbasis accident, other than a SGTR, is within the accident ana1-ysis assumptions as described in thet App-Li-_a __S~a fty-An-lyses secti n, The acci ent indu-ce-d leakage rate inclu sannprimary to seconda-r LEAKAGE existingprior to the ac _dent in aidtion to tsLjaKAGE i;Ldu/pd d-urin.nq the agac~ident-The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit ono*pe onaL__LEA is nt ' in LCO 3.4.6.-2, "RCS Opearational LEAKAGE," and limits rimary to secon-daryLEAKAGE through any one SG eo15g0 _q ped _Tis__Jii__hmaAedon iss the as sumptinthata single crack leaking this amount would not propagate to a SGTR under the stress conditions of a _LOCA or bmainteamilineJbreak.

If this emount of LEAKAGE is due to more than one crack -the cracks are very Pma-l--__annd the above ass__uption is conservati APPLICABILITY Steam generator tube integrity is challenged when the pressure d ifferentkijaacross the tubeAsislargqe. Large differenti apr-ess-es across SG tubes can only be experienced in MODE !2. 3. or 4.

I Providedfor Information Only.

C__i *n_*__faeless chan1gin MqDqianMODED_5___an-d__6 thandmrin-q MODES 1, 2, 3, and 4, In MODES 5 and 6,-primarv to secondary differential pressure is low, resul*In in lower stresses and r*d_*ed ptentaL for-LEAKAGE.

ACTIQNS T*heeACTIONS ar-e Mo-ieaNotedclkafr-iv thathe aetidnaybe entered indepe ndntlyfor each SG tube hisi accepble be s he require d actions provide ap~p roriate compnsatorv -actions for for continued opreration, and subsequently affected SG tubes are qoverned bys*_sentn coniti *onentrvpyli pli~cnoZia *_o iaJ**ed reciu red a jin s a, ACTION aapplies if itis isovered-hat one or more SG tubes exami ned in an inserv inspectiýon atisf _he tb reai_ criteria but were .not Dlugged or repaired in accordance with the Steaminxrator Procgram as requireddb SR 4.4.5.1. Aney*_axitinof SG t-ubeintegrity of-the affected tube(s) must be made, Steam generator tube int] agrit onxeetinog the SG pDefonrance crgi~tei.a desc bled in the St*am Genera 3L Pizro m The SGrepair criteria _dfme limits on SG- bu derada tion that-allow for flaw girowwth between inspections whie stj-ill providhing assurance that-the SG0Performance criteria will continue to be met. In ore od__rz~e_*_*SG_~eta hux have been oilUqqed or repaired has tube interi tv. an 9_valD~illDDdL~e___o*plJ~__t1aJ__dg~~ts tatthe SG performance criteria will continue to be met until the next ref _eJinouxagqe or SG tube inspection. The tube inte determination is baed on theestimated condition of the tvube at the time the situation is discovered and the estimated-growth of the degqrdaij-iprior to the next SG tube inspection, If it is det ermine dthat t-iube integrit isnobin mlna~inieid. Act ion__b__ppJ~ijeS_

Ac-oipletion time of 7 afiu fficient to compete the evaluation while minimizing the risk of pjlant operation with a SG tubethat may not have tubeinteqrit If the evaluation determines that the affected tbe(s) have b ~elgrity, q ACTION a41 ws__pi Wq to continue until the next refueling outaoe or SG inspection provided the inspection interval continues to be supported bvan operational assessment that reflects the affected tubes, However* -the a ff ected tube(s) must be1plu-qedor repaired p-rior to entering MODE 4followjnq thenext refu4eling outage or SG inspection. This completion time is

-a-ceptable since operation-until..the next inspection is sup~prted by the operationa.l assessment.

b If the reguired _tions and associatedcompletion times of ACTION a are not met or if SG tube inteqrity is not being maintained, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the followin- 30 houxi-s

I Providedfor Information Only.

The allowed com1 etion t im1esarere*__ nabelb1, Iaed___on operating eperince, to reach the desired plant cndio insi_

from f ~luower conditions in an orderly manner and without chaIIe ngingq p-_amtnsstms S-URVYJAT KEUIRME=I SR 4.4.5.1 Durin shudown periods the SGs are inspeced asr__gxiired b v thisSR a the Stm Generator_ M. NE1 97-06, "Steam Generator PrograM guidelines" and its referenced EPRI Guidelines establish the content Df the Steam GeneratorPrograim. Use of the Steam Generator Program ensures that the inspection is apporoxpriate an* consistent with D!,ringSG inspections a condit in onitoringy.asessment of the SG tubes is pgerformed, The condition monitorijgassessment determines

=the "as found" condition of the SGtubes, The purpose of the

. t, moni-to~rin-q__asea t o en sux-e__hh eS G__p rfo~rmaaa_

criteria have been met for the Dreviousperatingaperiod_

The Ste m Generator Proqcrrm in conjunction with the deqxradation

.ssessment determines the scope of the inspectionnand the methods usedto determine whether the tubes conta*n flaws satisfvin-qthe tube repair criteria, Inspection scope (i.e., which tubes or areas of

-tvbjq within the SG are to be inpctedisa function of existing an potential deqradation locations. The Steam Generator Program and theg__ertion ant also speji ythe inspection methods to be used to find potential deqradation. Inspection methods are a

-uInction of dear~adaion mopiIky odsr/7tve exa-mi-na-tion-IND-E-_

_echniquuecpabiI ites_ andinsect ion loations The St-e-amGenera rqmr-agrdmefines the Fr-enency of SR 4.4.5. . The Frequency is determined by the operational assessment and other Simi.ts in EPRI. "Pressurized Water Reactor Steam Generator ixamination Ginideelins." The Steam Generator information on existin rowth rates to determine jnspection Frequency that provides reasonable assurance that .the tubing will meet the SG performance criteria at the next scheduled inspection. In _fication 6.19 conteciptv reguirements concerningjnspection intervals to provide aded rance that the SG exrformac c rj eriJawil be met between s-hhe~du Iedin spe c i ng SR 4.4.5.2 D-u-lnrq__an__SGi-Anspection. any-_in p-cted eth-_sU sf s-fi the Steam Generator Pro rm raaair criteiaisrepired or removed from seryice kyplqggin.r. The tue__j_*_~i**_g~~~!9di p.iicl 6.19 are intended to ensure that tubes accepted for continued service

~t~sfA __theSperfo e c a witihiJw ane for erroDr in the flaw size measurement and for future flaw growth, In addition. the tube repair criteria, in coniunction with other elements of the Steam e tor Program, ensure that the SG continue to be met until the next inspection of the subiect tube(s).

I Providedfor Information Only.

NEI 97-06 provides _qidance for performin operational assessments to verify that the tubes remaining in service will continue to meet the B*4pe~rioxmd~ance criteria.

Steamgqeneratortbeepairs are onlyperfo d sin __appxoved repair methods as described in the Steam Generator Pro ram The Freency of -prior to entering MODE 4 followin a SG inpection" ensures__that SR 4.4.5.2 h s bn mplete nd all tubes meetinqthe

_rpma r c rite ~aaregq __ r ixep iare-d px Qro to su -c-ti he SG

• _tbes to sinificant rimary to seconda~r v__ sre differentia.l BEAVER VALLEY - UNIT 2 B 3/4 4-3b Change No. a---I2-_03_i