ML20210B898

Final Typed Pages for LARs 109 & 115,previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115

ML20210B898
Person / Time
Site:	Beaver Valley
Issue date:	07/15/1999
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20210B896	List: L-99-111, Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes ML20210B898
References
NUDOCS 9907230269
Download: ML20210B898 (35)

Text

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.' ATTACHMENT A Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change Nos. 109 & 115 Applicable Typed Pages 1

i l

9907230269 993715 PDR ADOCK 05000412 P PDR

ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. NPF-73

, DOCKET NO. 50-412 Replace the following pages of Appendix A, Technical Specifications, with the. enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating ,

the areas of change. l

?

EamoXn Insert {

i 1-4 1-4 1-5 1-5 1-6 1-6 1-7 1-7 3/4 4-12 3/4 4-12 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4-14

~3/4 4-14a_ 3/4 4-14a 3/4 4-14b 3/4 4-14b 3/4 4-14c 3/4 4-14d 3/4 4-14e 3/4 4-14f 3/4 4-16 3/4 4-16 3/4 4-19 3/4 4-19

'3/4 4-27 3/4 4-27 l 3/4 4-28 3/4 4-28  :

l 3/4 4 3/4 4-29 B 3/4 4-3 B 3/4 4-3 B 3/4 4-3a B 3/4 4-3a B 3/4 4-3b B 3/4 4-4e B 3/4 4-4e B 3/4 4-4f B 3/4 4-4f B 3/4 4-4g B 3/4 4-4g ]

B 3/4 4-4h B 3/4 4-4h j B 3/4 4-41 B 3/4 4-41 i B 3/4 4-4j B 3/4 4-4j B 3/4 4-4k B 3/4 4-5 B 3/4_4-5 B 3/4 4-6 B 3/4 4-6 {

B,3/4 7-2j B 3/4 7-2j l I

E .

a

, NPF -

_ DEFINITIONS-l ' 1'.15 THROUGH'1.17 (DELETED)

OUADRANT POWER TILT RATIO (OPTR) l 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, whicheverLis greater.

l

! ' DOSE EOUIVALENT I-131 1.19' DOSE EQUIVALENT I-131 shall be that concentration of I-131

-(microcuries/ gram) that alone would produce the same thyroid dose as i the quantity and. isotopic mixture'of I-131, I-132, I-133, I-134, and I-135 ac.tually present. The DOSE EQUIVALENT I-131 is calculated with the.following equation:

I C g_134 + CI-135 Cg_131 + C -132 + C -133 +

I

=

Cg_131 D .P

• 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;.

b. The testing of' one - (1) system, subsystem, train or other designated component at the beginning of each subinterval.

FREOUENCY-NOTATION'

1.21 The FREQUENCY NOTATION specified for the performance of Surveillance . Requirements shall correspond to the intervals defined in Table'1.2. j REACTOR' TRIP SYSTEM RESPONSE TIME i

. 1.22 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the tilde intervalLfrom when the monitored parameter exceeds its trip setpoint

'at the channel sensor until loss of stationary gripper coil voltage.

l-

~ BEAVER VALLEY - UNIT 2 1-4 Amendment No.

i I

L

e 4

NPF-73

- DEFINITIONS

! . ENGINEERED SAFETY FEATURE' RESPONSE TIME l'.23 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when' the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their. required positions, pump discharge pressures reach their l

required values, etc.). Times shall include diesel generator starting and sequence loading' delays where applicable.

- AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFERENCE shall be the difference in normalized i

flux signals between : the top - and bottom halves of a two-section excore neutron detector.

PHYSICS TESTS i

1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the' reactor core and.related instrumentation and 1) described in Chapter 14.0 of the FSAR,

2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

E - AVERAGE DISINTEGRATION ENERGY 1.26 E shall be the average sum (weighted in proportion to 'the concentration of each radionuclide in the reactor coolant at the time energies per i of samplir;} of the average beta and gamma l disintegration (in MeV) for. isotopes, other than iodines, with half l lives greater than 15 minutes, making up at least 95% of the total 4 L non-iodine activity in the coolant.

SOURCE CHECK

1.27 A SOURCE CHECK shall be the qualitative assessment of channel

response when the channel sensor is exposed to a radioactive source.

PROCESS CONTROL PROGRAM l-1.28 The' PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test', and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid

' wastes will be accomplished in such a way as to assure compliance I

with 10 CFR Parts 20, 61, and.71, State regulations, burial ground

-requirements, and other requirements-governing the disposal of solid .

i radioactive waste.

l' 1.29 DELETED BEAVER VALLEY - UNIT 2 1-5 Amendment No.

l L

. NPF-73 DEFINITIONS OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.30'- . The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoj , s, and in the conduct of the , Environmental Radiological Monitoring Program. The ODCM shall also contain- (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.6 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by

. Specifications in the Administrative Control section.

GASEOUS RADWASTE TREATMENT SYSTEM 1.31 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting Primary Coolant System offgases from the primary system and providing for delay .or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

VENTILATION EXHAUST TREATMENT SYSTEM f

1.32 VENTILATION EXHAUST TREATMENT SYSTEM is any system designed <

and installed to reduce gaseous radioiodine or radioactive material l in particulate form in ' ef fluents by passing ventilation or vent i exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust streau prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric. cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintnin temperature, pressure, humidity, concentration or- other operating conditions, in such a manner that replacencnt air or gas- is required to purify the confinement.

VENTING 1.34 VENTING.is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or 'other operating conditions, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

BEAVER VALLEY - UNIT 2 1-6 Amendment No. j

. NUF-73 DEFINITIONS l

l MAJOR CHANGES 1.35 MAJOR CHANGES to radioactive waste systems (liquid, gaseous and solid), as addressed in the PROCESS CONTROL PROGRAM, shall include the following:

1) MAJOR CHANGES in process equipment, components, structures, and effluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) or the Hazards Summary Report and evaluated in the staff's Safety Evaluation Report (SER) (e.g., deletion of evaporators and installation of demineralizers; use of fluidized bed calciner/ incineration in place of cement solidification systems);

2) MAJOR CHANGES in the design of radwaste treatment systems (liquid, gaseous, and solid) that could significantly increase the quantities or activity of effluents released or volumes of solid waste stored or shipped offsite from those previously considered in the FSAR and SER (e.g.,

use of asphalt system in place of cement);

3) Changes in system design which may invalidate the accident analysis as described in the SER (e.g., changes in tank capacity that would alter the curies released);

and

4) Changes in system design that could potentially result in a significant increase in occupational exposure of operating personnel (e.g., use of temporary equipment without adequate shielding provisions). i l

MEMBER (S) OF THE PUBLIC 1.36 MEMBER (S) OF THE PUBLIC means any individual except when that individual is receiving an occupational dose.

CORE OPERATING LIMITS REPORT 1.37 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current i operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.5. Plant operation within these operating limits l is addressed in individual specifications.

i BEAVER VALLEY - UNIT 2 1-7 Amendment No.

NPF-73

, . REACTOR COOLANT-SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

1. 'All nonplugged tubes 'that previously had detectable wall penetrations greater than 20 percent, and

2. Tubes in those areas where experience has indicated potential problems, and

-3. At least 3 percent of the total _ number of sleeved tubes in.all'three steam gtnerators. A sample size less than - 3 percent le acceptable provided all the sleeved tubes in the steam generator (s) examined during the refueling outage are inspected. These inspections will include both the tube and the sleeve, and

4. A tube inspection pursuant to Specification 4.4.5.4.a.S. If any selected tube does not permit the passage of the eddy current probe for a tube or sleeve i inspection, this shall be recorded and an adjacent

!- tube shall be selected and subjected to a tube inspection.

i

5. Indications left in service P.s a result of application of the tube support plate voltage-based repair criteria (4.4.5.4.a.10) shall be inspected by bobbin coil probe during all future refueling outages.

c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection l may.be subjected to a partial tube inspection provided:

l l- _ 1. - The tubes selected for these samples include the tubes l from those areas of the tube sheet array where tubes with imperfections were previously found, and l

l 2. The inspections include those portions of the tubes where imperfections were previously found.

d. Implementation of the steam generator tube-to-tube support i 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 diarater 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 thei) full length.

BEAVER VALLEY - UNIT 2 3/4 4-12 Amendment No.

I

I NPF-73

. EEACTOR COOLANT SYSTEM

~ SURVEILLANCE REQUIREMENTS (Continued)

The results of each sample inspection shall be classified into one of the following three categories:

Cateaory Inspection Results C-1 Less than 5 percent of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1 percent of the total tubes inspected are defective, or between 5 percent and 10 percent of the total tubes inspected are i

degraded tubes.

C-3 More than 10 percent of the total I

tubes inspected are degraded tubes or more than 1 percent of the inspected tubes are defective.

, Note: In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10 percent) further wall

. penetrations to be included in the above percentage calculations.

l 4.4.5.3 Ingoection Frecuencies - The above required inservice l

inspections of steam generator tubes shall be performed at the following frequencies: j l

a. The first inservice inspection shall be performed after  !

6 Effective Full Power Months but within 24 calendar months I

. of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.

If two consecutive inspections following servive under All Volatile Treatment (AVT) conditions, not including the l preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections- demonstrate that previously observed degradation has not continued and no additional degradation ,

has occurred, the inspection interval may be extended to a l maximum of once per 40 aonths.

i i

BEAVER VALLEY - UNIT 2 3/4 4-13 Amendment No.

l

f- ,

4 HPF-73 3'2 ACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

=

b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection .whose results fall in Category C-3, the l

inspection frequency shall be increased to at least once l per 20 months. The increase in inspection frequency shall l apply until a subsequent inspection demonstrates that a third sample inspection is not required.

i c. Additional, unscheduled inservice inspections shall be

! performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1. Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2,

2. A seismic occurrence-greater than the Operating Basis Earthquake,

3. A loss-of-coolant accident requiring actuation of the engineered safeguards, or

4. A main steamline or feedwater line break.

4.4.5.4 Acceotance Criteria

a. As used in this Specification:

1. Imoerfection means an exception to the dimensions, finish or contour of _ a tube or sleeve from that required by fabrication drawings or specifications.

Eddy-current testing indications below 20 percent of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2.

Dearadation means a service-induced cracking,

wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve.

3. peoraded Tube means a tube or sleeve containing imperfections greater'than or equal to 20 percent of the nominal wall thickness caused by degradation.

l BEAVER VALLEY - UNIT 2 3/4 4-14 Amendment No.

~

l l J

. i

,' NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

4. Percent Dearadation means the percentage of the tube or sleeve wall thickness affected or removed by degradation.

5. Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube containing a defect is defective. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.

6. Pluccina or Reoair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the affected area because it may become unserviceable prior to the next inspection. The plugging or repair limit imperfection depths are specified in percentage of nominal wall thickness as follows:

a) Original tube wall 40% l This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to 4.4.5.4.a.10 for the repair limit applicable to these intersections.

b) ABB Combustion Engineering TIG welded 32% l sleeve wall c) Westinghouse laser welded sleeve wall 25% l

7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its I structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steamline or feedwater line break as specified in 4.4.5.3.c, above.

8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot-leg side) l completely around the U-bend to the top support to the cold-leg. l

-BEAVER VALLEY - UNIT 2 3/4 4-14a Amendment No.

~

NPF-73

- REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

9. Tube 'Renair refers to sleeving which is ussd to maintain a tube in-service or return a tube to service. This includes the removal of plugs that were installed as a corrective or preventive measure. The following sleeve designs have been found actepu.ble:

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

b) Westinghouse laser welded sleeves, WCAP-13483, Revision 1.

-10. Tube Suncort Plate Pluacina Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predon!nantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. ,At tube support plate intersections, the plugging- (repair) limit is based on maintaining steam generator tube serviceability as described below; a) Steam generator tubes, whose degradation is 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, whose degradation is 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 4.4.5.4.a.10.c below.

4 i

BEAVER VALLEY - UNIT 2 3/4 4-14b Amendment No.

NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 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 limitill may remain in service if ' _a rotating pancake coil or 4 acceptable alternative inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater j than the upper voltage repair limitill will be l plugged or repaired.

d). If- an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits )

apply instead of the limits identified in 4.4.5.4.a.10.a, 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.

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

SL YHURL ~

/CL - A t' 1.0 + NDE + Gr

' ~

\ CL >

1. CL - 8t' l Vggg =Vggg - ( Vgg - Vgg)

(1) The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.

t

-BEAVER. VALLEY - UNIT 2 3/4 4-14c Amendment No. l

~ '

NPF ,, REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) where:

4 Vme = upper voltage repair limit j Vuc = lower voltage repair

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

Vu22 = mid-cycle lower voltage l

repair limit based on Vams and time into cycle At = length of time since last scheduled inspection during which Vue and Vuc were implemented CL = cycle length (the time I

between two scheduled steam generator inspections)

Vet = 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)

Implementation of these mid-cycle repair limits should follow ' the same approach as in TS 4.4.5.4.a.10.a,

[ 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.

(2). The NDE is the value provided by the NRC in GL 95-05 as supplemented.

BEAVER VALLEY.- UNIT 2 3/4 4-14d Amendment No. l

1

, NPF-73

, REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging or repair limit) required by Table 4.4-2.

4.4.5.5 Reports

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be submitted in a Special Report in accordance with 10 CFR 50.4.

b. The complete results of the steam generator tube and sleeve inservice inspection shall be submitted in a Special Report in accordance with 10 CFR 50.4 within 12 months following the completion of the inspection. This Special Report shall include:

1. Number and extent of tubes and sleeves inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged or repaired.

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to Specification 6.6 prior to resumption of plcnt operation. The written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

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

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

BEAVER VALLEY - UNIT 2 3/4 4-14e Amendment No. l l 1

. NPF-73

,. REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

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

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

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

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

l l

l BEAVER VALLEY - UNIT 2 3/4 4-14f Amendment No. l

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1 e i NPF-73

.. REACTOR COOLANT SYSTEM OP'ERATI'ONAL 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.

I ACTION:

a. With any pressure boundary LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any Reactor Coolant Sy8 am LEAKAGE greater than any one of the above limits, excluding pressure boundary LEAKAGE, reduce the LEAKAGE rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6' hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s: W

1. Containment atmosphere gaseous radioactivity monitor.

l

-(1) Only on leakage detection instrumentation required by LCO  ;

3.4.6.1.

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

l 1

j

. NPF-73

,. REACTOR COOLANT SYSTEM

' ^

3 /4. 4.'8 SPECIFIC' ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. s 0.35 pCi/ gram DOSE EQUIVALENT I-131, and l

b. 5 100/E pCi/ gram APPLICABILITY: . MODES 1, 2, 3, 4, and 5 ACTION:

MODES 1, 2 and 3*:

a. , With the specific activity of the primary coolant > 0.35 l pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

-during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with Tavg

< 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the. specific activity of the primary coolant > 100/E pCi/ gram, -be in HOT STANDBY with Tavg < 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

MODES 1, 2, 3, 4, and 5 a.- With the specific activity of the primary coolant > 0. 3 5 ' l pCi/ gram DOSE EQUIVALENT I-131 or > 100/E pCi/ gram, perform ,

t the sampling analysis requirement of item 4a of Table 4.4-12 until the specific activity of the primary coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS

'4.4.8 The specific activity of the primary coolant shall be determined to . be within the performance limits of the sampling and analysis program of Table 4.4-12.

• With T.vg- 2 500*F. - .

BEAVER VALLEY - UNIT 2 3/4 4-27 Amendment No.

[

, NPF-73

. . TABLE 4.4-12 j I

l PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE I AND ANALYSIS PROGRAM TYPE OF MEASUREMENT MINIMUM- MODBS IN WHICH AND ANALYSIS FREOUENCY SURVEILLANCE REOUIRED

1. Gross Activity 3 times per 7 days 1, 2, 3, 4 Determination with a maximum time l of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between l l samples.

. :2. Isotopic Analysis 1 per 14 days 1, l for DOSE EQUIVALENT l I-131 Concentration

3. Radiochemical for 1 per 6 months 1, E Determination

4. Isotopic Analysis a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1#,2#,3#,4#,5#

for Iodine whenever the including I-131 specific activity I-133, and I-135 exceeds 0.35 I pCi/ gram DOSE EQUIVALENT I-131 or 100/E pCi/ gram, and b) One sample between 1, 2, 3 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a 1-hour period.

l

1. Until the specific activity of the primary coolant system is restored to within its limits.

BEAVER VALLEY - UNIT 2 3/4 4-28 Amendment No.

l i

L

NPF-73 1

e 250 \

a 5

'b 2

]

D 5: 200 8

g N UNACCEPTABLE OPEF:ATION E

D 150 5

8-O 2 100 E

a 2 ACCEPTABLE OPERATION

% 50 \

8 w

8 0

20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.35 pCl/ gram Dose Eq'ulvalent 1-131 BEAVER VALLEY - UNIT 2 3/4 4-29 Amendment No.

l l

. NPF-73.

,- REACTOR COOLANT SYSTEM BASES  ;

3/4.4.5 STEAM GENERATORS (Continued) decay heat removal capabilities for RCS temperatures greater than 350'F if one steam generator becomes inoperable due to single failure .

considerations. Below 350 F, decay heat is removed by the RHR I system.

l The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision . 1. Inservice inspection of steam generator tubing is I essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means or characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corresion cracking. The extent of cracking during plant operation would te limited by the limitation of steam generator tube leakage between the Primary Coolant System and the Secondary Coolant System (prinary-to-secondary LEAKAGE = 150 gallons per day per steam generator). Axial cracks having a primary-to-secondary LEAKAGE less than this limit during operation will have an adequate margin of safety to withstand the loads imposer! during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary LEAKAGE of 150 gallons per day per steam generctor can readily be detected. LEAKAGE in excess of this limit will require plant shutdown and .n unscheduled inspection, during which the leaking tubes will te .ocated and plugged or repaired by sleeving. The technical bases for sleeving are described in the approved vendor reports listed in Surveillance Requirement 4.4.5.4.a.9.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in serv 5ce, it will be found during scheduled inservice steam generator tube examinations. Plugging or repair will be required of all tubes with imperfections exceeding the plugging or repair limit. Degraded steam generator tubes may be repaired by the installation of sleeves which span the degraded tube section. A steam generator tube with a sleeve installed meets the structural BEAVER VALLEY - UNIT 2 B 3/4 4-3 Amendment No.

,' NPF-73

,- REACTOR COOLANT SYSTEM BASES 2/4.4.5 STEAM GENERATORS (Continued) requirements of tuber, which are not degraded, therefore, the sleeve is considered a part. of the tube. The surveillance equirements identify those sleeving methodologies approved for use. If an installed nieeve is found to have through wall penetration greater than or equ al to the plugging limit, the tube must be plugged. The plugging li: nit for the sleeve is derived from R. G. 1.121 analysis which utilir,es a 20 percent allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional degradation growth. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20 percent of the original tube wall thickness.

The voltage-based repair limits of these surveillance requirements (SR) implement the guidance in GL 95-05 and are applicable only to Westinghouse-designed steam generators (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections. The guidance in GL 95-05 will not be applied to the tube-to-flow distribution baffle plate intersections. The voltage-based repair limits are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG. Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with no NDE detectable cracks extending  ;

outside the thickness of the support plate. Refer to GL 95-05 for additional description of the degradation morphology.

Implementation of these SRs requires a derivation of the voltage j structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).

The voltage structural limit is the voltage from the burst prea.ture/ bobbin voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 650'F (i.e., the 95-percent LTL curve). The voltage structural limit must be adjusted downward to account for potential degradation growth during an operating interval and to account for NDE uncertainty. The upper voltage repair limit; Vugn, is determined from the structural voltage limit by applying the following equation:

VURL = Vst - VGr - ViiDE BEAVER VALLEY - UNIT 2 B 3/4 4-3a Amendment No.

6

, NPF-73 t

,- REACTOR COOLANT SYSTEM BASES M ,4.5 STEAM GENERATORS (Continued) where var represents the allowance for degradation growth between inspections and Vuos represents the ellowance for potential sources of error in the measurement of the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.

Safety analyses were performed pursuant to Generic Letter 95-05 to determine the maximum MSLB-induced primary-to-secondary leak rate that could occur without offsite doses exceeding a small fraction of 10 CFR 100 (concurrent iodine spike), 10 CFR 100 (pre-accident iodine spike), and without control room doses exceeding CDC-19. The current value of the maximum MSLB-induced leak rate and a summary of the

)

analyses are provided in Section 15.1.5 of the UFSAR.

The mid-cycle equation in SR 4.4.5.4.a.10.d should only be used during unplanned inspections in which eddy current data is acquired for indica'; ions at the tube support plates.

1 SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 for situations which the NRC wants to be notified prior to returning the SGs to service. For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle (EOC) voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EO voltage distributions prior to returning the SGs to service. Nota that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.1 and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should be provided per the GL section 6.b (c) criteria.

Wher.ever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation. Such cases will be considered by the commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

BEAVER VALLEY - UNIT 2 B 3/4 4-3b Aldendment No. l

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) ,

i BACKGROUND (Continued) l During plant life, the joint and valve interfaces can produce varying I amounts of reactor coolant LEAKAGE, through either normal operational I 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 j LCO specifies the types and amounts of LEAKAGE. z 1

10 CFR 50, Appendix A, GDC 30, requires means for detecting and, to i 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 l source, rate, and duration. Therefore, detecting and monitoring I 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 I 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 j boundary (RCPB) from degradation and the core from inadequate /l cooling, in addition to preventing the accident analyses radiation l release assumptions from being exceeded. The consequences of  !

violating this LCO include the possibility of a loss of coolsnt i accident (LOCA).  !

APPLICABLE SAFETY ANALYSES Except for primary-to-secondary LEAKAGE, the safety analyses do not l address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect t.he probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes a 1 gpm 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.

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

2.

m ,

L

. NPF-73

,- ' REACTOR COOLANT SYSTEM i

BASES-3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

APPLICABLE SAFETY ANALYSES (Continued)

-Primary-to-secondary LEAKAGE is a factor.in the dose releases outside containment resulting from a MSLB accident. To a lesser extent, l other accidents or transients. involve secondary steam release to the atmosphere,- such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The MSLB is more limiting for site radiation releases. The primary-to-secondary ; LEAKAGE assumed in the safety analysis for the . MSLB accident- is ' described in UFSAR Section 15.1.5. The radiological consequences of : . a MSLB outside of containment was reanalyzed in j support of .the. tube support plate voltage-based repair . criteria l stated in . SR : 4. 4. 5. 4. a .10. For this analysis, the thyraid dose was maximized at 10% of the 10 CFR Part 100 guideline of 300 rem for the

'co-incident iodine spike case. RCS leakage was based on projection rather than on technical specification leakage limits. The analysis indicated that-offsite doses-would remain within regulatory criteria with the assumed primary-to-secondary leakage (described in UFSAR Section 15.1.5) should. steam generator tubes fail due to the depressurization associated with a MSLB.

A similar analysis was performed using a control room thyroid dose of

30 rem as the criterion. The control room was assumed to be manually

' isolated and pressurized at T=30 minutes for a period of one hour, at which time filtered emergency intake would be automatically started.

The control room would be purged wid fresh air at T=8 hours following release cessation. The analysis indicated that control room dases would remain within regulatory criteria with the' assumed primary-to-secondary leakage -(described in UFSAR Section 15.1.5) should steam generator tubes fail due to the depressurization associated with a MSLB.

.LCQ 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 tyDe is unacceptable as the leak itself could cause turther deterioration,- resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB.

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

NPF-73

, REACTOR COOLANT SYSTEM 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. Primarv-to-Secondary LEAKAGE throuch Any One SG Operating experience at PWR plants has shown that sudden >

increases in leak rate are often precursors to larger tube failures. Maintaining an operating LEAKAGE limit of l 150 gpd per steam generator will minimize the potential for j a large LEAKAGE event at power. This operating LEAKAGE 1 limit is more restrictive than the operating LEAKAGE limit in standardized technical specifications. This provides

{

additional margin to accommodate a tube flaw which might grow at a greater than expected rate or unexpectedly extend outside the thickness of the tube support plate. This  !

reduced LEAKAGE limit, in conjunction with a leak rate {

monitoring program, provides additional assurance that this precursor LEAKAGE will be detected and the plant shut f.own in a timely manner.

I l

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

m .

[ NPF-73

. REACTOR COOLANT SYSTEM l

L BASES l 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

LCO (Centinued)

d. Identified LEAKAGE l

Up to 10 Jpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere

.with detection of identified LEAKAGE and is well within the l 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).

i 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.

L LCO 3.4.6.2, "RCS Pressure Isolation Valve (PIV)," measures leakage through each individual PIV and can impact this LCO, Of t!1a two PIVs in series in each isolated line, leakage measurec' 'hrough one PIV.

L does not result in RCS LEAKAGE when the other is leak tight. If both I 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 Amendment No.

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

ACTIONS

g. If any pressure bcundary LEAKAGE exists, 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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.

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

If the unidentified LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action reduces the LEAKAGE.

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 4-41 Amendment No.

l I

q NPF-73 EEACTOR COOLANT SYSTEM  !

BASES 3/4~4.6.2 OPERATIONAL LEAEAGE (Continuerl*

SURVEILLANCE REOUIREMENTS ISR)

SR 4.4.6.2 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 perfomance of an RCS water i inventory balance. Primary to secondary LEAKAGE is also measured by j performance of an RCS water inventory balance in conjunction with j effluent monitoring within the secondary steam and feedwater systems.

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

Therefore, this SR is not required to be performed in MODES 3 and 4 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation near operating pressure have a been established.

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

For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and  ;

RCP seal injection and return flows.

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

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency is a reasonable interval to trend LEAKAGE and j recognizes the importance of early leakage detection in the !

prevention of accidents. Note (1) states that the 12 hour I surveillance is required only on leakage detection instrumentation l required by LCO 3.4.6.1. This Note allows the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> monitoring to be suspended on leakage detection inctrumentation which is inoperable or not required to be operable per LCO 3.4.6.1. Note (2) states that this SR is required to be performed during steady state operation. j I

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

NPF-73

-REACTOR COOLANT SYSTEM

' BASES 3/4.4.6.3 PRESSURE ISO QTION VALVE LEAKAGE Tne 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 isole. tion valve is identified LEAKAGE and will be considered as a portion of the allowed limit.

B'EAVER' VALLEY - UNIT'2 B 3/4 4-4k Amendment No. 1

. 1

)

NPF-73 En CTOR COOLANT SYSTEM i l

l BASES 3/4.4.7 CHEMISTRY l

The limitations on Reactor Coolant System chemistry ensure that j corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.

The associated ef.Cects of exceeding the oxygen, chloride and fluoride 1 limits are time and temperature dependent. Corrosion studies show  !

that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for ]

j the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to-take corrective action.

3/4.4.8 C SPECIFIC A.C_TIV_IlX l The primary coolant specific activity is limited in order to maintain

-offsite and control room operator doses associated with postulated

accidents within applicable requirements. Specifically, the 0.35 l pCi/gm DOSE EQUIVALENT I-131 limit ensures that the offsite dose does i not exceed a small fraction of 10 CFR Part 100 guidelines and that control ' room operator thyroid dose does not exceed GDC-19 in the event of primary-to-secondary leakage induced by a main steam line break.

The ACTION statement permitting POWER OPERATION to continue for l l

limited time periods with the primary coolant's specific activity '

> L.35 pCi/ gram, DOSE' EQUIVALENT I-131, but within the allowable limit l shown on Figure 3.4-1, accommodates possible iodine spiking i phenomenon. which may occur following changes in THERMAL POWER. l Operation with specific activity' levels excee ding 0.35 pCi/ gram DOSE l l EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or ' exceeding the limits shown on Figure 3.4-1 must be j restricted to ensure that assumptions made in the UFSAR accident  !

analyses are not exceeded.

Reducing Tavg to < 500*F minimizes the release of activity should a l j

' steam generator tube rupture since the saturation pressure of the i primary coolant is below the lift pressure of the atmospheric steam i

BEAVER VALLEY - UNIT 2 B 3/4 4-5 Amendment No.

t

9 ,

. NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.8 SPECIFIC ACTIVITY (Continued)

, relief valves. This action also reduces the pressure differential across the steam generator tubes reducing the probability and magnitude of main steam line break accident induced primary-to-secondary leakage. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Ccolant System are designed to withstand the effects of cyclic loada due to system temperature and pressure changes. These cyclic loadr,. are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various' categories of load cycles used for design purposes are provided .in Section 3.9 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal-induced compressive i stresses tend to alleviate the tensile stresses induced by the internal pressurt Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermi stresses) represents a l lower bound of all similar curves for fi3 ite heatup rates when the l inner wall of the vessel is treated as the governing location.  !

The heatup analysis also covers the determination of pressure- I temperature.limitatinns for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients

- established during heatup produce tensile stresses at the outer . fall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced  ;

stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup  :

ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

BEAVER VALLEY - UNIT 2 B 3/4 4-6 Amendment No.

I

1 l NPF-73 PLANT SYSTEMS 1

BASES 3/4.7.1.3 PRIMARY PLANT DEMINERALIZED WATER (?PDW)

The OPERABILITY of the PPDW storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours with steam discharge to atmosphere.

I 3/4.7.1.4 ACTIVITY '

The limitations on secondary system specific activity ensure that steam releases to the environment will not be significant contributors to radioactivity releases resulting from analyzed accidents. Many of the analyzed accidents assume that a loss of auxiliary AC power occurs, making the main condenser unavailable for plant cooldown, and making it necessary to dump steam to the environment via SG atmospheric dump valves. Maintaining secondary system specific activity within the limits ensures that these releases, in conjunction with other releases associated with the accident, will be within applicable dose criteria.

l l

l BEAVER VALLEY - UNIT 2 B 3/4 7-2j Amendment No.

l l

i ATTACHMENT B Beaver Valley Power Station, Unit No. 2 Prcposed Technical Specification Change Nos. 109 & 115 l -

e .

Change Summary Typed Pages Editorial Chances 1-4 No additional changes are included.

1-5 1-6 1-7 In Definition 1.37, for the Core Operating Limits Report, reference to Specification "6.9.1.12" is replaced with Specification "6.9.5" since the specification number was changed in Amendment 97.

3/4 4-12 No additional changes are included.

3/4 4-13 s 3/4 4-14 Surveillance Requirement 4 4.5.3.c.4 was marked-up to replace " steam line" with "steamline"; however, this change was already incomorated by Amendment 98 so this change is not required.

3/4 4-14a Surveillance Requirement 4.4.5.4.a.7 was marked-up to i replace " steam line" with "steamline"; however, this change was already incorporated by Amendment 98 so this change is not required.

3/4 4-14b No additional changes are included.

3/4 4-14c l 3/4 4-14d 3/4 4-14e i 3/4 4-14f l 3/4 4-16 3/4 4-19 3/4 4-27 3/4 4-28 3/4 4-29 B 3/4 4-3 B 3/4 4-3a B 3/4 4-3b l

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P.

- Attachment B i

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Page 2 B 3/4 4-4e The sentence proposed to be added to this paragraph has been modified to clarify the exception used for the .mam steamline break analysis to support the voltage-based tube repair criteria. The modified sentence states: "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."

B 3/4 4-4f The first paragraph has been modified by replacing " steam line break (SLB)" with "MSLB" since main steamline break (MSLB) is defined on the previous page and to be consistent with the terminology used in the next paragraph. The second and third paragraphs have been revised to incorporate the same changes addressed for Unit 1. This was provided by Letter L-98-193 dated Septemter 30,1998, where the second and last sentences in the second paragrah and in the last sentence of the third paragraph were modified by replacing the specific primary-to-secondary leakage value with reference to UFSAR Section 15.1.5. This will avoid inconsistencies between the UFSAR and the technical specification Bases.

B 3/4 4-4g The changes proposed in LAR 109 were superseded by the B 3/4 4-4h changes in LAR 115.

B 3/4 4-4i No additional changes are included.

B 3/4 4-4j i B 3/4 4-4k B 3/4 4-5 B 3/4 4  !

B 3/4 7-2j i e

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