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St. Lucie Unit 1 L-2010-078 Docket No. 50-335 St. Lucie Unit 1 EPU LAR Att. 4-1 ATTACHMENT 4 LICENSE AMENDMENT REQUEST EXTENDED POWER UPRATE TECHNICAL SPECIFICATIONS BASES MARKUPS (For Information Only)
FLORIDA POWER & LIGHT ST. LUCIE NUCLEAR PLANT UNIT 1 This coversheet plus 24 pages Technical Specifications Bases Markups
St.. Lucie Unit 1 L-2010-078 Docket No. 50-335 LIST OF PAGES St. Lucie Unit 1 EPU LAR Att. 4-2 Technical Specifications Bases Markups Technical Specification Bases Section 2.0 3 of 10 4 of 10 8 of 10 Section 3/4.1 5 of 9 6 of 9 Section 3/4.2 5 of 5 Section 3/4.4 7 of 29 21 of 29 INSERT 1 22 of 29 23 of 29 24 of 29 28 of 29 Section 3/4.5 4 of 6 Section 3/4.6 3 of 10 4 of 10 6 of 10 Section 3/4.7 3 of 13 4 of 13 5 of 14 Section 3/4.9 2 of 9 9 of 9 Section 3/4.11 3 of 3
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The trip setpoint is bounding relative to the accident and transient analyses which were performed using a lower, conservative trip setpoint.
The trip setpoint and the methodology used to determine the trip setpoint, the as-found acceptance criteria band, and the as-left acceptance criteria are specified in the UFSAR. The two footnotes on the bottom of TS Table 2.2-1 are consistent with the two recommended notes provided in NRC's letter to the NEI Technical Setpoint Methods Task Force for Setpoint Allowables dated September 7, 2005.
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70f29 SECTION NO.:
3/4.4 REVISION NO.:
3 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 6 OF ADM-25.04 REACTOR COOLANT SYSTEM S1. LUCIE UNIT 1 PAGE:
I I
l 3/4.4 3/4.4.5 REACTOR COOLANT SYSTEM (continued)
, and 2) the pre-existing secondary BASES (continued) side fluid inventory
~
STEAM GENERATORS (SG) TUBE I~
II"I"ntriinued
-,activity in the d
The analysis for design basis acciden s and transients ther an a SG TR assume the SG tubes retain their stru,~tural integrity (i.
., they are assumed not to rupture). In these analyses the steam discharge to the atmosphl re is It
'1) based o'1t.the total primary-to-secondary leakage from all SGs of 1 gpm and I wo sources.
0.5 gpm through anyone SG as a result of accident induced condition~V
---__----, For accidents that do not involve fuel damage, the primary coolant activity
," and the secondary level of DO~~ ~QUI\\l/\\L~~JT I 1d1 is assumed to be equal to the limits in coolant system activity is LCO 3.4.8, "Reactor Coolant System Specific Activity"; For accidents that assumed to be equal to assume fuel damage, the primary coolant activity is a function of the amount the limits in LCO of activity released from the damaged fuel. The dose consequences of 3.7.1.4, "Plant Systems these events are within the limits of GDC 19 (Ref. 2), 10 G~R 100 (Ref. d)
Activity."
or tRe ~~RG approved liseRsiR§ Basis (e.§., a sfflall frostioR of tRese lifflits). ~
L....------r------l Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(C)(2)(ii)\\
Limiting Condition for Operation (LCO) land 10 CFR 50.67 (Ref. 7).
The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.
During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.
A SG tube has tube integrity when it satisfies the SG performance criteria.
The SG performance criteria are defined in Specification 6.8.4.1, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.
There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. Failure to meet anyone of these criteria is considered failure to meet the LCO.
SECTION NO.:
TITLE:
TECHNICAL SPECIFICATIONS PAGE:
I 3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 21 of 29 REVISION NO.:
S1. LUCIE UNIT 1 I
3/4.4 REACTOR COOLANT SYSTEM (continued) l BASES (continued) 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce 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 effects of exceeding the oxygen, chloride and fluoride 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 the specified limit 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.
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~~RG sf tYJ=)isal site IssatisRs. TAoso >rall:los aro sSRsor>rati>ro iR tAat sJ=)osifis site J=)ararTIotors sf tAO ~t. Ll:lsio sito, SI:lSA as sito eSI:lRsary IssatisR aRs rTIotosrels§isal sSRsitisRs, wore RSt sSRsisores iR tAis ovall:latisR.
TAo /\\GTIO~~ statOrTIORt J=)orrTIittiR§ j:20W~R OJ:2~R/\\TIO~~ ts sSRtiRl:lo fsr lirTIitos tirTIo J=)orisss witA tAO J=)rirTIary ssslaRt's sJ=)osifis asti>rity > 1.0
~Git§rarTI DO~~ ~QUIVl\\L~~n I 1d1, el:lt witAiR tAO allswaelo lirTIit SASWR SR
~i§l:lro d.4 1, asssrTlrTlssatos J=)sssielo issiRo sJ=)il(iR§ J=)AORSrTlORSR wAisA rTIay sssl:lr fslIswiR§ oAaR§os iR TI=IERM/\\L j:20WER
INSERT 1 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY The maximum allowable doses to an individual at the exclusion area boundary (EAB) distance for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an accident, or at the low population zone (LPZ) outer boundary distance for the radiological release duration, are specified in 10 CFR 50.67 for design basis accidents using the alternative source term methodology and in Branch Technical Position 11-5 for the waste gas decay tank rupture accident. Dose limits to control room operators are given in 10 CFR 50.67 and in GDC 19.
The RCS specific activity LCO limits the allowable concentration of radionuclides in the reactor coolant to ensure that the dose consequences of limiting accidents do not exceed appropriate regulatory offsite and control room dose acceptance criteria. The LCO contains specific activity limits for both DOSE EQUIVALENT (DE) I-131 and DOSE EQUIVALENT (DE) XE-133.
The radiological dose assessments assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator tube leakage rate at the applicable Technical specification limit. The radiological dose assessments assume the specific activity of the secondary coolant is at its limit as specified in LCO 3.7.1.4, Plant Systems - Activity.
The ACTIONS allow operation when DOSE EQUIVALENT I-131 is greater than 1.0 Ci/gram and less than 60 Ci/gram. The ACTIONS require sampling within four hours and every four hours following to establish a trend.
One surveillance requires the determination of the DE XE-133 specific activity as a measure of noble gas specific activity of the reactor coolant at least once per 7 days.
A second surveillance is performed to ensure that iodine specific activity remains within the LCO limit once per 14 days during normal operation and following rapid power changes when iodine spiking is more apt to occur. The frequency between two and six hours after a power change of greater than 15% RATED THERMAL POWER within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established because the iodine levels peak during this time following iodine spike initiation.
The RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
SECTION NO.:
3/4.4 REVISION NO.:
3 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 6 OF ADM-25.04 REACTOR COOLANT SYSTEM S1. LUCIE UNIT 1 PAGE:
I 22 of 29 I
3/4.4 REACTOR COOLANT SYSTEM (continued)
BASES (continued) 3/4.4.RSPE"",C.....I~=+I....
C----+t~t.C::r-THI¥'J+-IITHYH(l-Eai9oRlntEtfi ntt:l:Ite~s) l 3/4.4.9 Resl:lain§ Tavg to < eOO°F" prevents tAe release of aativity SAol:lls a stearn
§enerator tl:lbe rl:lptl:lre sinae tAe satl:lration pressl:lre of tAe primary ooolant is below tAe lift pressl:Ire of tAe atmospAerio stearn relief valves.
TAe sl:IF'v'eillanae re~l:Iirements provise ase~l:Iate assl:Iranae tAat o>wessive speaifia aativity levels in tAe primary aoolant will be seteotes in sl:Iffioient time to talm oorreotion aotion. Information obtaines on iosine spil~in§ will be l:Ises to assess tAe parameters assoaiates witA spil~in§ pAenomena. /\\
resl:lation in fre~l:Ienay of isotopia analyses followin§ power aAan§es may be permissible if jl:lstifies by tAe sata obtaines.
PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads 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 5.2.1 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 through the reactor vessel wall produce thermal stresses which are compressive at the reactor vessel inside surface and are tensile at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside surface and outside surface locations, the total applied stress is greatest at the outside surface location. However, since neutron irradiation damage is larger at the inside surface location than at the outside surface location, the inside surface flaw may be more limiting.
Consequently, for the heatup analysis, both the inside surface and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting.
During cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and are compressive at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface.
SECTION NO.:
3/4.4 REVISION NO.:
3 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 6 OF ADM-25.04 REACTOR COOLANT SYSTEM S1. LUCIE UNIT 1 PAGE:
I 23 of 29 I
3/4.4 REACTOR COOLANT SYSTEM (continued)
BASES (continued) l 3/4.4.9 PRESSURIZER/
TEMPERATURE LIMITS (continued)
Since neutron irradiation damage is also greater at the inside surface, the inside surface flaw location is the limiting location during cooldown.
Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.
70 The heatup and cooldc wn limit curves (Figures 3.4-2a and 3.4-2b) are composite curves which were prepared by determining the most conservative case, wit either the inside or outside wall controlling, for the heatup rate of up to eQ~F/hr and for any cooldown rate of up to 100°F per hour. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the applicable service period.
The reactor vessel materials have been tested to determine their initial RTNOT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E>1 Mev) irradiation will cause an increase in the RTNOT. Therefore, an adjusted reference temperature can be calculated based upon the fluence. The heatup and cooldown limit curves shown on Figures 3.4-2a and 3.4-2b include predicted adjustments for this shift in RTNOT at the end of the applicable service period, as well as adjustments for pressure differences between the reactor vessel beltline and pressurizer instrument taps.
The actual shift in RTNOT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-82, reactor vessel material surveillance specimens installed near the inside wall of the reactor vessel in the core area. The capsules are scheduled for removal at times that correspond to key accumulated fluence levels within the vessel through the end of life. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, measured.'1RTNOT for surveillance samples can be applied with confidence to the corresponding material in the reactor vessel wall. The heatup and cooldown curves must be recalculated when the
.'1RTNOT determined from the surveillance capsule is different from the calculated.'1RTNOT for the equivalent capsule radiation exposure.
The pressure-temperature limit lines shown on Figures 3.4-2a and 3.4-2b for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements for Appendix G to 10 CFR 50.
SECTION NO.:
3/4.4 REVISION NO.:
3 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 6 OF ADM-25.04 REACTOR COOLANT SYSTEM S1. LUCIE UNIT 1 PAGE:
I 24 of 29 I
3/4.4 REACTOR COOLANT SYSTEM (continued)
BASES (continued) l 3/4.4.9 PRESSURIZER/
TEMPERATURE LIMITS (continued)
This Lowest Service Temperature value of 165°F also includes an additional rF to account for temperature measurement uncertainty.
The maximum RTNOT for all reactor coolant system pressure-retaining materials, with the e;ception of the reactor pressure V~c:::c:::~1 h::l~ h~~ro 158 I
established to be ~r.
I ne Lowest Service Temperature limit line shown
'-----r----'
on Figures 3.4-2a and 3.4-2b is based upon this RTNOT since Article NB-2332 of Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNOT + 100°F for piping, pumps and valve~ Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia.
The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection program for ASME Code Class 1,2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. This program is in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a (g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).
Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.
SECTION NO.:
3/4.4 REVISION NO.:
3 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 6 OF ADM-25.04 REACTOR COOLANT SYSTEM S1. LUCIE UNIT 1 PAGE:
I 28 of 29 I
3/4.4 REACTOR COOLANT SYSTEM (continued)
BASES (continued) 3/4.4.11 DELETED 3/4.4.12 PORV BLOCK VALVES The opening of the Power Operating Relief Valves fulfills no safety related function. The electronic controls of the PORVs must be maintained OPERABLE to ensure satisfaction of Specifications 3.4.12 and 3.4.13.
Since it is impractical and undesirable to actually open the PORVs to demonstrate reclosing, it becomes necessary to verify operability of the PORV Block Valves to ensure the capability to isolate a malfunctioning PORV.
l 3/4.4.13 3/4.4.14 300 POWER OPERATED RELIEF VALVES and REACTOR COOLANT PUMP - STARTING The low temperature overpressure protection system (LTOP) is designed to prevent RCS overpressurization above the 10 CFR 50 Appendix G operating limit curves (Figures 3.4-2a and 3.4-2b) at RCS temperatures at or below of dblriR§ Reatblp aRd 2g1 of dblriR§ oooldowR. The LTOP syste IS based on the use of the pressurizer power-operated relief val s (PORVs) and the implementation of administrative and operational ntrols.
The PORVs aligned to the RCS with the low pressure setpoints of 350 and 530 psia, restrictions on RCP starts, limitations on heatup and cooldown rates, and disabling of non-essential components provide assurance that Appendix G PIT limits will not be exceeded during normal operation or design basis overpressurization events due to mass or energy addition to the RCS. The LTOP system APPLICABILITY, ACTIONS, and SURVEILLANCE REQUIREMENTS are consistent with the resolution of Generic Issue 94, "Additional Low-Temperature Overpressure Protection for Light-Water Reactors," pursuant to Generic Letter 90-06.
SECTION NO.:
3/4.5 REVISION NO.:
2 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 7 OF ADM-25.04 EMERGENCY CORE COOLING SYSTEMS (ECCS)
S1. LUCIE UNIT 1 PAGE:
40f6 I
I 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)
BASES (continued) 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS l
The OPERABILITY of two separate and independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the safety injection tanks is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.
TS 3.5.2.c and 3.5.3.a require that ECCS subsystem(s) have an independent OPERABLE flow path capable of automatically transferring suction to the containment sump on a Recirculation Actuation Signal.
The containment sump is defined as the area of containment below the minimum flood level in the vicinity of the containment sump strainers.
Therefore, the LCOs are satisfied when an independent OPERABLE flow path to the containment sump strainer is available.
- S 3.5.2, ACTION a.1. provides an allowed outage/action completion time (AOT) of up to 7 days from initial discovery of failure to meet the LCO provided the affected ECCS subsystem is inoperable only because its associated LPSI train is inoperable. This 7 day AOT is based on the findings of a deterministic and probabilistic safety analysis and is referred to as a "risk-informed" AOT extension. Entry into this ACTION requires that a risk assessment be performed in accordance with the Configuration Risk Management Program (CRMP) which is described in the Administrative Procedure (ADM-17.0B) that implements the Maintenance Rule pursuant to 10 CFR 50.65.
The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that at a minimum, the assumptions used in the accident analyses are met and that subsystem OPERABILITY is maintained.
TS 3.5.2.d requires that an ECCS subsystem(s) have OPERABLE charging pump and associated flow path from the BAMT(s). Reference to TS 3.1.2.2 requires that the Train A charging pump flowpath is from the BAMT(s) through 1....-----1the boric acid makeup pump(s). The Train B charging pump flowpath is from the BAMT(s) through the gravity feed valve(s).
SECTION NO.:
3/4.6 REVISION NO.:
5 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 8 OF ADM-25.04 CONTAINMENT SYSTEMS S1. LUCIE UNIT 1 BASES FOR SECTION 3/4.6 PAGE:
I 3 of 10 I
l 3/4.6 CONTAINMENT BASES 3/4.6.1 CONTAINMENT SYSTEMS VESSEL 3/4.6.1.1 CONTAINMEN T VESSEL INTEGRITY CONTAINMENT VESSEL INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
In accordance with Generic Letter 91-08, "Removal of Component Lists from Technical Specifications," the opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and this action will prevent the release of radioactivity outside the containment.
3/4.6.1.2 CONTAI NMENT LEAKAGE 42.771 The limitations on containment leakage rates ensure that to containment leakage volume will not exceed the value a med in the accident analyses at the peak accident pressure, Pa psig) which results from the limiting design basis loss of coolant accident.
The surveillance testing for measuring leakage rates is performed in accordance with the Containment Leakage Rate Testing Program and is consistent with the requirements of Appendix "J" of 10 CFR 50, Option B and Regulatory Guide 1.163 Rev. 0, as modified by approved exemptions.
3/4.6.1.3 CONTAI NMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.
SECTION NO.:
3/4.6 REVISION NO.:
5 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 8 OF ADM-25.04 CONTAINMENT SYSTEMS S1. LUCIE UNIT 1 PAGE:
4 of 10 3/4.6 CONTAINMENT SYSTEMS (continued)
BASES (continued) 3/4.6.1 CONTAINMENT VESSEL (continued) 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structural is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.70 psi and 2) the containment peak pressure does not exceed the design pressure of 44 psig during steam line break accident conditions.
0.5
~
3.08 The maximum peak pressure obt.
a steam line break accident is psig. The limit of pSlg for initial positive containment pressure will limit the total pressure to 44.0 psig which is the design pressure and is consistent with the accident analyses.
3/4.6.1.5 AIR TEMPERATURE The limitation on containment air temperature ensures that the containment vessel temperature does not exceed the design temperature of 264°F during LOCA conditions. The containment temperature limit is consistent with the accident analyses.
3/4.6.1.6 CONTAI NMENT VESSEL STRUCTURAL INTEGRITY The limitation ensures that the structural integrity of the ontainment steel vessel will be maintained comparable to the original sign standards for the life of the facility. Structural integrity is required 0 ensure that the vessel will withstand the maximum pressure of psig in the event of the limiting design basis loss of coolant accident. A visual inspection in accordance with the Containment Leakage Rate Testing Program is sufficient to demonstrate this capability.
SECTION NO.:
3/4.6 REVISION NO.:
5 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 8 OF ADM-25.04 CONTAINMENT SYSTEMS S1. LUCIE UNIT 1 PAGE:
I 6 of 10 I
3/4.6 CONTAINMENT SYSTEMS (continued) l 3/4.6.2 BASES (continued)
DEPRESSURIZATION AND COOLING SYSTEMS (continued) 3/4.6.2.1 CONTAI NMENT SPRAY AND COOLING SYSTEMS (continued)
Ensuring that the containment spray pump discharge pressure is met satisfies the periodic surveillance requirement to detect gross degradation caused by impeller structural damage or other hydraulic component problems. Along with this requirement,Section XI of the ASME Code verifies the pump developed head at one point on the pump characteristic curve to verify both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the unit safety analysis. Surveillance Requirements are specified in the Inservice Testing Program, which encompassesSection XI of the ASME Code.Section XI of the ASME Code provides the activities and frequencies necessary to satisfy the requirements.
3/4.6.2.2 SPRAY ADDITIVE SYSTEM
~ 17.0 I The OPERABILITY of the spray additive system ensures that sufficient
~
.NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between and 11.0 for the solution recirculated within containment after a LOCA.
This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. These assumptions are consistent with the iodine removal efficiency assumed in the accident analyses.
3/4.6.2.3 DELETED 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. This includes the containment purge inlet and outlet valves.
SECTION NO.:
3/4.7 REVISION NO.:
2 3/4.7 PLANT TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 9 OF ADM-25.04 PLANT SYSTEMS S1. LUCIE UNIT 1 BASES FOR SECTION 3/4.7 SYSTEMS PAGE:
I 3 of 13 I
l BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% of its design pressure of 1000 psia during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition and ASME Code for Pumps and Valves, Class II. The total relieving capacity for all valves on all of the steam lines is12.38 x 106 Ibs/hr wAioA is 102.8 r:>eFGeRt tAe total seooRsary stealTl flow of 12.04 )( 4e2 Ibs/Ar at 100% R/\\T~D TI=I~RM/\\L POW~R A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases:
'For two loop operation SP (X)-(Y)(V) x (1065)
X where:
SP =reduced reactor trip setpoint in percent of RATED THERMAL POWER V = maximum number of inoperable safety valves per steam line
SECTION NO.:
3/4.7 REVISION NO.:
2 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 9 OF ADM-25.04 PLANT SYSTEMS S1. LUCIE UNIT 1 PAGE:
I 4 of 13 I
3/4.7 PLANT SYSTEMS (continued) l BASES (continued) 3/4.7.1 TURBINE CYCLE (continued) 106.5 = Power Level-High Trip Setpoint for two loop operation x
= Total relieving capacity of all safety valves per steam line in Ibs/hour (6.192 x 106 Ibs/hr.)
+2/-3%
Y
= Maximum relieving capacity of anyone safety valve in Ibs/hour (7.74 x 105 Ibs/hr.)
Surveillance Requirement 4.7.1.1 verifies the OPERABILITY of e
1+/- 3% ~S~Vs by the verification of each MSSV lift setpoint in acc ance with
~~~ervice Testing Program. The MSSV setpoints a 1000 psia
.-1 r 1<, (4 valves each header) and 1040 psia
-1/ ')
(4 valves each header) for OPERABILITY; however, the valves are reset to 1000 psia
+/-1 % and 1040 psia +/- 1%, respectively, during the Surveillance to allow for drift. The LCO is expressed in units of psig for consistency with implementing procedures.
The provisions of Specification 3.0.4 do not apply. This allows entry into and operation in MODE 3 prior to performing the Surveillance Requirements so that the MSSVs may be tested under hot conditions.
3/4.7.1.2 AUXILIARY FEEDWATER PUMPS The OPERABILITY of the auxiliary feedwater pumps ensures that the Reactor Coolant System can be cooled down to less than 325°F from normal operating conditions in the event of a total loss of off-site power.
NSATESTORAGETANKS Any two of the three auxiliary feedwater pumps have the required capacity to provide sufficient feedwater flow to remove reactor decay heat and reduce the RCS temperature to 325°F where the shutdown cooling system may be placed into operation for continued cooldowr*rt-o-m-ai-n-ta-in-H-O-T-....o.,
STANDBY for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and then 3/4.7.1.3 CONDE The OPERABILITY of the condensate storage tank wit the minimum water volume ensures that sufficient water is available cooldown of the Reactor Coolant System to less than 325°F in the event of a4effil. loss of off-site power. The minimum water volume is sufficient to maintain the RCS at HOT STANDBY conditions for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with steam discharge to atmosphere. ~
The minimum usable volume to satisfy the criteria stated
~ above is 130,500 gallons, which is ensured by the LCO for the CST volume of 153,400 gallons.
SECTION NO.:
3/4.7 REVISION NO.:
4 TITLE:
TECHNICAL SPECIFICATIONS BASES ATTACHMENT 9 OF ADM-25.04 PLANT SYSTEMS S1. LUCIE UNIT 2 PAGE:
5 of 14
- 0*
\\.
3/4.7 3/4.7.1 PLANT SYSTEMS (continued)
BASES (continued)
TURBINE CYCLE (continued) 3/4.7.1.3 CONDE NSATE STORAGE TANKS The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the Unit 2 RCS at HOT STANDBY conditions for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> followed by an orderly cooldown to the shutdown cooling entry temperature (350°F). The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
1130,500~
The actual water requirements are 149,600 gallons for Unit 2 and gallons for Unit 1. Included in the required volumes of water are the tank unusable volume of gallons and a ~e allowance for instrument error of 21~ gallon.
9203 I
an 3/4.7.1.4 ACTIVITY 4230 I The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will comply with the dose criterion provided in 10 CFR 50.67 in the event of a steam line rupture. The dose also includes the effects of a coincident 1.0 gpm primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the safety analyses.
~
3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Requirements is consistent with the assumptions used in the safety analyses.
The specified 6.75 second full closure time represents the addition of the maximum allowable instrument response time of 1.15 seconds and the maximum allowable valve stroke time of 5.6 seconds. These maximum allowable values should not be exceeded because they represent the design basis values for the plant.
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gaseous radioactive waste inventory in a effective dose equivalent 1
Branch Technical Position 11-5, "Postulated Radioactive Releases Due to Waste Gas System Leak or Failure," of Standard Review Plan Chapter 11, Radioactive Waste Management," of NUREG-0800.
The waste gas decay tank inventory source term required to generate an exclusion area boundary dose of 0.1 rem is the basis for the limit of 202,500 dose equivalent curies Xe-133, and is derived based on the definition given in Technical Specification Task Force (TSTF)-490, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.