ML20195K182
| ML20195K182 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/15/1999 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20195K180 | List: |
| References | |
| NUDOCS 9906220020 | |
| Download: ML20195K182 (25) | |
Text
{{#Wiki_filter:[. . ATTACHMENT 109 Technical Specification Paces s I k Il 1 f
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9906220020 990615 PDR ADOCK 05000412 P ppg ,
ATTACHMENT TO LICENSE AMENDMENT NO. , FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 l I 1 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. i l Remove Insert I 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 4-26 3/4 4-16 3/4 4-19 3/4 4-19 B 3/4 4-3 B 3/4 4-3 B 3/4 4-3a B 3,'4 4-3a B 3/s 3b 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 B 3/4 4-41 B 3/4 4-41 j B 3/4 4-4j B 3/4 4-4j j B 3/4 4-4k j 1
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'APF-73 REACTOR COOLANT SYSTEM i
SURVEILLANCE REQUIREMENTS (Continued) i
- 1. All nonplugged tubes that previously had detectable l l
wall penetrations greater than 20 percent, and
- 2. Tubes in those areas where experience has indicated l potential problems, and l l
- 3. At least 3 percent of the total number of sleeved tubes in all three steam generators. A sample size less than 3 percent is acceptable provided all the sleeved tubes in the steam generator (s) examined l 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.8. If any selected tube does not permit the i 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. 5, Indications left in service as 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 may be subjected to a partial tube inspection provided:
- 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
- 2. The inspections include those portions of the tubes where imperfections were previously found,
- d. 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 l
indications shall be based on the performance of at least a 20-percent random sampling of tubes inspected over their full length. l BEAVER VALLEY - UNIT 2 3/4 4-1L Amendment No. L
. 'NPF-7'3 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l The results of each sample inspection shall be classified into one of l the following three categories:
i Catecory 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 j degraded tubes.
)
C-3 More than 10 percent of the total tubes inspected are degraded tubes or more than 1 percent of l 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. 4.4.5.3 Inspection Frecuengles - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection shall be performed after j 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections l 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 service under All Volatile Treatment (AVT) conditions, not including the j preservice inspection, result in all inspection results I 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 , maximum of once per 40 months. { l l l BEAVER VALLEY - UNIT 2 3/4 4-13 Amendment No. 2
' PF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 1
- 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 inspection frequency shall be increased to at least once l per 20 months. The increase in inspection frequency shall apply until a subsequent inspection demonstrates that a third sample inspection is not required.
- 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 Acceptance Criteria
- a. As used in this Specification:
- 1. Imperfection 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. Dearaded Tube means a tube or sleeve containing imperfections greater than or equal to 20 percent of the nominal wall thickness caused by degradation.
BEAVER VALLEY - UNIT 2 3/4 4-14 Amendment No.
'EPF-73 REACTOR COOLANT SYSTEM SURVEIIO."CR DFCL REMENTS (Continued)
- 4. Percent Decradation 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 Repair Limit means the imperfection depth at or beyond which the tube shall be remova 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 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. I BEAVER VALLEY - UNIT 2 3/4 4-14a Amendment No.
- l
~ 'NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
- 9. Tube Repair refers to sleeving which is used to maintain a tube in-service or return a tube to service. This includes the removal of plugs that were t installed as a corrective or preventive measure. The I following sleeve designs have been found acceptable:
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. l l
- 10. T_ube Succort Plate Pluccina Limit is used for the {
disposition of an alloy 600 steam generator tube for ' continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below: a) Steam generator tubes, 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. i f l f i l l l l ( BEAVER VALLEY - UNIT 2 3/4 4-14b Amendment No. ) I i
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'NPF-73 REACTOR COOLANT SYSTEM i l
SURVEILLANCE REQUIREMENTS (Continued) c) Steam generator tubes, with indications of i potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube sr.pporc plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (1) may remain in service if a rotating pancake coil or acceptable alternative inspection does not detect ' degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin volta greater than the upper voltage repair limit (1)ge will be j plugged or repaired. 1 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 1 following equations: VsL Vgggt = 1.0 + NDE + Gr A CL > i i ICL - dth ) Vgtat =V MURL - (Vggt -V tat)\ CL /
)
l (1) The upper voltage repair limit is calc.ulated according to the l methodology in Generic Letter 95-05 as supplemented. l l l BEAVER VALLEY - UNIT 2 3/4 4-14c Amendment No. l
F
. 'NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) where:
Vme = upper voltage repair limit Vuc = lower voltage repair limit Vmmt = mid-cycle upper voltage repair limit based on l time into cycle j Vmat = mid-cycle lower voltage repair limit based on Vmmt and time into cycle At = length of time since last scheduled inspection during which Vem and Vum were implemented CL = cycle length (the time between two scheduled steam generator inspections) Vst = 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, i 4.4.5.4.a.10.b, and 4.4.5.4.a.10.c.
l l l i (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
4 , .
' PF-7'3 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 plant 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 criccria 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 ond-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.
BEAVER VALLEY - UNIT 2 3/4 4-14e Amendment No. l
- PF-7'3 BEBCTOR 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.
- 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~ , notify the Commission and provide an assessment of the safety significance of the occurrence.
BEAVER VALLEY - UNIT 2 3/4 4-14f Amendment No. l
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'NPF-7'3 i REACTOR COOLANT SYSTEM !
OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION c 3.4.6.2 Reactor Coolant System operational LEAKAGE shall be limited to* I I
- a. No pressure boundary LEAKAGE,
- b. 1 gpm unidentified LEAKAGE, !
i
- c. 150 gallons per day primary-to-secondary LEAKAGE through any one steam generator, and
- d. 10 gpm identified LEAKAGE.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With any pressure boundary LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the next 30 hours.
- b. With any Reactor Coolant System LEAKAGE greater than any one of the above limits, excluding pressure boundary LEAKAGE, reduce the LEAKAGE rate to within limits within 4 hours or be in at least HOT STANDBY within the next ,
6 hours and in COLD SHUTDOWN within the following 30 hours. 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 l j at least once per 12 hours:(1)
- 1. Containment atmosphere gaseous radioactivity monitor. I l
i I (1)_ Only on LEAKAGE detection instrumentation required by LCO l 3.4.6.1. J l s BEAVER VALLEY - UNIT 2 3/4 4-19 Amendment No.
- PF-73 REACTOR COOLANT SYSTEM l BASES I
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 I corciderations. Below 350 F, decay heat is removed by the RHR ' system. 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 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 of 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 corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube LEAKAGE between the Primary Coolant System and the Secondary Coolant System (primary-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 imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary LEAKAGE of 150 gallons per day per steam generator can readily be detected. LEAKAGE in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located 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 service, 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-7'3 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continuedi requirements of tubes which are not degraded, therefore, the sleeve is considered a part of the tube. The surveillance requirements identify those sleeving methodologies approved for use. If an installed sleeve is found to have through wall penetration greater than or equal to the plugging limit, the tube must be plugged. The plugging limit for the sleeve is derived from R. G. 1.121 analysis which utilizes 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 cegradation morphology.
Implementation of these SRs requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation 1 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 pressure / 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; VURL, is determined from the structural voltage limit by applying the following equation: VURL " VsL - Vor - VNDE BEAVER VALLEY - UNIT 2 B 3/4 4-3a Amendment No.
. 'NPF-7'3 l REACTOR COOLANT SYSTFM BASES 3/4.4.5 STEAM GENERATORS (Continued) i 1
l where Vor represents the allowance for degradation growth between l inspections and Vuos represents the allowance for potential sources i of error in the measurement of the bobbin coil voltage. Further l discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05. , l The mid-cycle equation in SR 4.4.5.4.a.10.d should only be used i during unplanned inspections in which eddy current data is acquired l for indications at the tube support plates. l l 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 i requirement, LEAKAGE and conditional burst probability can be l calculated based on the as-found voltage distribution rather than the ! projected end-of-cycle (EOC) voltage distribution (refer to GL 95-05 i for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to i returning the SGs to service. Note that if LEAKAGE and conditional f burst probability were calculated using the measured EOC voltage l 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. Whenever 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, i laboratory examinations, tests, additional eddy-current inspection, j and revision of the Technical Specifications, if necessary. i I l I l BEAVER VALLEY - UNIT 2 B 3/4 4-3b Amendment No. l 2
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'NPF-7'3 REACTOR COOLANT SYSTFE l 1
BASES J 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) i BACKGROIJND (Continued) j During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE ) from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE. 10 CFR 50, Appendix A, GDC 30, requires means for detecting and, to I the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 describes acceptable methods for selecting LEAKAGE. detection systems. l The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the ) operators, allowing them to take corrective action should a leak l occur that is detrimental to the safety of the facility and the l public. l 1 l 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 atmosphnre, if possible, to not interfere with RCS LEAKAGE detection. l This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA). APPLICABLE SAFETY ANALYSES Except for primary-to-secondary LEAKAGE, the safety analyses do not l address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of LEAKAGE can l affect the 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. BEAVER VALLEY - UNIT 2 B 3/4 4-4e Amendment No.
'NPF-73 REACTOR COOLANT SYSTEM BASES 2]4.4.6.2 OPERATIONAL LEAKAGE (Continuedl APPLICABLE SAFETY ANALYSES (Continued)
Primary-to-secondary LEAKAGE is a factor in the dose releases outside containment resulting from a main steamline break (MSLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The LEAKAGE contaminates the secondary fluid. l The MSLB is more limiting for site radiation releases. The primary-to-secondary LEAKAGE assumc5 in the safety analysis for the MSLB accident is described in UFSAR Section 15.1.5. The radiological consequences of a MSLB outside of containment was reanalyzed in support of the tube support plate voltage-based repair criteria , stated in SR 4. 4. 5. 4.a.10. For this analysis, the thyroid dose was maximized at 10% of the 10 CFR Part 100 guideline of 300 rem for the co-incident iodine spike case. RCS LEAKAGE was based on projection rather than on technical specification LEAKAGE limits. The analysis indicated that offsite doses would remain within regulatory criteria with the assumed primary-to-secondary LEAKAGE (described in UFSAR Section 15.1.5) should steam generator tubes fail due to the depressurization associated with a MSLB. A similar analysis was performed using a control room thyroid dose of 30 rem as the criterion. The control room was assumed to be manually isolated and pressurized at T=30 minutes for a period of one hour, at which time filtered emergency intake would be automatically started. The control room would be purged with fresh air at T=8 hours following release cessation. The analysis indicated that control room doses would remain within regulatory criteria with the assumed primary-to-secondary LEAKAGE (described in UFSAR Section 15.1.5) should steam generator tubes fail due to the depressurization associated with a MSLB. C LCO_ RCS operational LEAKAGE shall be limited to:
- a. Pressure Boundarv LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB.
LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Should pressure boundary LEAKAGE occur through a BEAVER VALLEY - UNIT 2 B 3/4 4-4f Amendment No.
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-Secondarv LEAKAGE thrcuch Any One SG Maintaining an operating LEAKAGE limit of 150 gpd per steam generator will minimize the potential for a large LEAKAGE event during a MSLB. Based on the non-destructive examination uncertainties, bobbin coil voltage distribution, and crack growth rate from the previous inspection, the expected leak rate following a steamline rupture is limited to the assumed primary-to-secondary LEAKAGE (described in UFSAR Section 15.1.5) in the faulted loop. Maintaining LEAKAGE less than or equal to the assumed primary-to-secondary LEAKAGE (described in UFSAR Section 15.1. 5) limit will ensure that postulated offsite doses will remain within the 10 CFR 100 requirements and that control room habitability continues to meet GDC-19.
LEAKAGE in the intact loops will be limited to the operating limit of 150 gpd. If the projected end-of-cycle distribution of crack indications results in primary-to-secondary LEAKAGE greater than the assumed primary-to-secondary LEAKAGE (described in UFSAR Section 15.1.5) in the faulted loop during a postulated steamline break event, additional tubes must be removed from service or repaired in order to reduce the postulated steamline break LEAKAGE to less than or equal to the assumed primary-to-secondary LEAKAGE (described in UFSAR Section 15.1.5). BEAVER VALLEY - UNIT 2 B 3/4 4-4g Amendment No.
NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) LCO (Continued) Also, the 150 gallons per day LEAKAGE limit incorporated into this specification is more restrictive than the standard operating LEAKAGE limit and is intended to provide an additional margin to accommodate a crack which might grow at a greater than expected rate or unexpectedly extend outside the thickness of the tube support plate. Hence, the reduced LEAKAGE limit, when combined with an effective leak rate monitoring program, provides additional assurance that should a significant leak be experienced in service, it will be detected, and the plant shut down in a timely manner,
- d. Identified LEAKAGE l Up to 10 gpm 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 capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE).
Violation of this LCO could result in continued degradation of a component or system. APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized. In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE. LCO 3.4.6.2, "RCS Pressure Isolation Valve (PIV)," measures LEAKAGE l through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, LEAKAGE measured through one PIV l does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE. BEAVER VALLEY - UNIT 2 B 3/4 4-4h Amendment No.
NPF-73 REACTOR COOLANT SYSTEM BASES 2/4.4.6.2 OPERATIONAL LEAKAGE (Continued) ACTIONS
- a. If any pressure boundary 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 hours and MODE 5 within 36 hours. 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.
- b. Unidentified LEAKAGE, identified LEAKAGE, or primary-to- l secondary LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours. This Completion Time allows time to verify LEAKAGE rates and either l 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 l limits within 4 hours, 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 hours and MODE 5 within 36 hours. 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.
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NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued) SERVEILLANCE REOUIREMENTS (SR) 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 performance of an RCS water inventory balance. Primary-to-secondary LEAKAGE is also measured by l performance of an RCS water inventory balance in conjunction with effluent monitoring within the secondary steam and feedwater systems. The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure. Therefore, this SR is not required to be performed in MODES 3 and 4 until 12 hours of steady state operation near operating pressure have been established. Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and a Note requires the Survoillance to be met when steady state is established. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor the containment atmosphere radioactivity and the containment sump level. The 12 hour 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." The 72 hour Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early LEAKAGE detection in the l prevention of accidents. Note (1) states that the 12 hour surveillance is required only on LEAKAGE detection instrumentation l required by LCO 3.4.6.1. This Note allows the 12 hour monitoring to be suspended on LEAKAGE detection instrumentation which is inoperable l 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. BEAVER VALLEY - UNIT 2 B 3/4 4-4j Amendment No. l l
NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.6.3 PRESSURE ISOLATION VALVE LEAKAGE The LEAKAGE from any RCS pressure isolation valve is sufficiently low l to ensure early detection of possible in-series valve failure. It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. LEAKAGE from the RCS pressure isolation valve is identified LEAKAGE l and will be considered as a portion of the allowed limit. BEAVER VALLEY - UNIT 2 B 3/4 4-4k Amendment No. l I I y . ---
ATTAClIMENT B-109 Chance Summary Tyned Pages Editorial Changes 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 incorporated 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 replace " steam line" with "steamline"; however, this change was already incorporated by Amendment 98 so this change is not required. 3/4 4-14e Surveillance Requirement 4.4.5.5.d.1 has been modified by replacing " leakage" with " LEAKAGE" since this is a defined term. 3/4 4-19 Surveillance Requirement 4.4.6.2.a and Note (1) have been modified by replacing " leakage" with " LEAKAGE" since this is a defined term. B 3/4 4-3 The third paragraph has been modified by replacing " leakage" with " LEAKAGE" since this is a defined tenn. B 3/4 4-3b The third paragraph has been modified by replacing " leakage" with " LEAKAGE" since this is a dermed term. B 3/4 4-4e The second, fourth and sixth paragraphs have been modified by replacing " leakage" with " LEAKAGE" since this is a defined tenn. The first paragraph under APPLICABLE SAFETY ANALYSES has been revised by replacing
" primary to secondary" with " primary-to-secondary" to be consistent with this convention used elsewhere. In addition, the sentence proposed to be added to this paragraph has been modified to clarify the exception used for the main steamline break analysis to support the voltage-based tube repair criteria. The modified sentence states:"An exception to the primety-to-secondary LEAKAGE is described below for the main steamline break (MSLB) analyzed in support of voltage-based steam generator tube repair criteria."
l 1 l
Attachment B-109 Change Summary, Unit 2 Page 2 B 3/4 4-4f The first paragraph has been modified by replacing "(SLB)" with "(MSLB)" to be consistent with terminology used elsewhere. The second and third paragraphs were revised to incorporate the same changes addressed for Unit 1 (per letter L-98-193 dated September 30,1998) by replacing the specific primary-to-secondary leakage value with reference to the applicable UFSAR section. This will avoid inconsistencies between the UFSAR and the technical specification Bases. The first, second and third paragraphs have also been modified by replacing " leakage" with " LEAKAGE" since this is a defined term. B 3/4 *-4g The first paragraph under Primary-to-Secondary LEAKAGE Throuch Any One SG has been revised by replacing" main steamline break" with "MSLB" and the proposed specific primary-to secondary leakage value has been replaced with reference to the applicable UFSAR section. This will avoid inconsistencies between the UFSAR and the technical specification Bases and is consistent with the above change. B 3/4 4-4h The first and last paragraphs have been modified by replacing
" leakage" with " LEAKAGE" since this is a defined term.
B 3/4 4-4i The Bases for Action b has been revised by replacing
" primary to secondary" with "primany-to-secondary" to be consistent with this convention used elsewhere. The third paragraph has been modified by replacing " leakage" with
" LEAKAGE" since this is a defined term.
B 3/4 4-4j The Bases for Surveillance Requirement 4.4.6.2 has been revised by replacing " primary to secondary" with " primary-to-secondaty" to be consistent with this convention used elsewhere. The Bases for Surveillance Requirement 4.4.6.2 has also been revised by replacing " leakage" with
" LEAKAGE" since this is a defined term.
B 3/4 4-4k The Bases for PRESSURE ISOLATION VALVE LEAKAGE has been revised by replacing " leakage" with " LEAKAGE" since this is a defined term. ,}}