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{{Adams
#REDIRECT [[ONS-2015-006, Technical Specification (TS) Bases Change]]
| number = ML15035A549
| issue date = 01/29/2015
| title = Oconee Nuclear Station - Technical Specification (TS) Bases Change
| author name = Batson S L
| author affiliation = Duke Energy Carolinas, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000269, 05000270, 05000287
| license number =
| contact person =
| case reference number = ONS-2015-006
| document type = Letter, Technical Specification, Bases Change
| page count = 112
}}
 
=Text=
{{#Wiki_filter:DUKE Scott L. BatsonVice President ENE LRGYIOconee Nuclear StationDuke EnergyONOIVP 1 7800 Rochester HwySeneca, SC 2967210 CFR 50.36 o: 864.873.3274 f 864.873.4208 ONS-2015-006 Scott.Batson@duke-energy.com January 29, 2015ATTN: Document Control DeskU.S. Nuclear Regulatory Commission 11555 Rockville PikeRockville,.
Maryland 20852
 
==Subject:==
 
Duke Energy Carolinas, LLCOconee Nuclear StationDocket Numbers 50-269, 50-270, and 50-287Technical Specification (TS) Bases ChangePlease find attached changes to the Oconee Nuclear Station (ONS) TS Bases. These changeswere processed in accordance with the provisions of Technical Specification 5.5.15, "Technical Specifications (TS) Bases Control Program."
TS Bases (TSB) Change 2014-05 revises TS Bases 3.4.3, RCS Pressure and Temperature Limits, Limited Condition for Operation section on the intent of the "any" period of time used inTS Tables 3.4.3-1, "Operational Requirements for Unit Heatup" and 3.4.3-2, "Operational Requirements for Unit Cooldown."
This clarification is needed to ensure that the heatup andcooldown rates are evaluated over a continuous period of time.Amendments 388/390/389 were issued to remove obsolete information associated with theReactor Protective System/Electrical System upgrades, Low Pressure Service WaterWaterhammer Prevention System modifications, and Emergency Condenser Circulating WaterSystem upgrades.
TS Bases (TSB) Change 2014-13 revises TS Bases 3.3.1, 3.3.3, 3.3.4, 3.3.5,3.3.6, 3.3.7, 3.3.27, 3.6.5 and 3.7.7 consistent with these amendments.
Any questions regarding this information should be directed to Sandra Severance, ONS Regulatory Affairs Group, at (864) 873-3466.
Sincerely, Scott L. BatsonVice President Oconee Nuclear StationAttachment Aw)I- (CCwww.duke-energy.com U. S. Nuclear Regulatory Commission January 29, 2015Page 2cc: Mr. Victor McCree, Regional Administrator U.S. Nuclear Regulatory Commission, Region IIMarquis One Tower245 Peachtree Center Ave., NE, Suite 1200Atlanta, GA 30303-1257 Mr. James R. Hall, Senior Project Manager (ONS)(By electronic mail only)U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 11555 Rockville PikeMail Stop O-8G9ARockville, MD 20852Mr. Jeffrey A. Whited, (Acting)
Project Manager (ONS)(By electronic mail only)U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 11555 Rockville PikeMail Stop O-8B1ARockville, MD 20852Mr. Eddy CroweSenior Resident Inspector Oconee Nuclear Station DUKE OconeeENERGY,7800SenJanuary 29, 2015Re: Oconee Nuclear StationTechnical Specification Bases ChangePlease replace the corresponding pages in your copy of the Oconee Technical Specifications Bases Manual as follows:luclear StationDuke EnergyRochester Hwyeca, SC 29672REMOVE THESE PAGESINSERT THESE PAGESList of Effective Pages (LOEP) 1- 17TS Bases Page B 3.4.3-1 thru 8TS Bases Page B 3.3.1-1 thru 30TS Bases Page B 3.3.3-1 thru 6TS Bases Page B 3.3.4-1 thru 4TS Bases Page B 3.3.5-1 thru 17TS Bases Page B 3.3.6-1 thru 4TS Bases Page B 3.3.7-1 thru 8TS Bases Page B 3.3.27-1 thru 6TS Bases Page B 3.6.5-1 thru 11TS Bases Page B 3.7.7-1 thru 6List of Effective Pages (LOEP) 1- 17TS Bases Page B 3.4.3-1 thru 8TS Bases Page B 3.3.1-1 thru 26TS Bases Page B 3.3.3-1 thru 4TS Bases Page B 3.3.4-1 thru 4TS Bases Page B 3.3.5-1 thru 15TS Bases Page B 3.3.6-1 thru 4TS Bases Page B 3.3.7-1 thru 8TS Bases Page B 3.3.27-1 thru 6TS Bases Page B 3.6.5-1 thru 10TS Bases Page B 3.7.7-1 thru 6If you have any questions concerning the contents of this Technical Specification Bases update,contact Sandra Severance (864) 873-3466.
C 4s WasikRegulatory Affairs Managerwww.duke-energy.com Attachment Oconee Nuclear StationRevised Technical Specification Bases Pages OCONEE NUCLEAR STATIONTECHNICAL SPECIFICATIONS-BASES REVISED 01/14/15LIST OF EFFECTIVE PAGESPAGE AMENDMENT BASES REVISION DATELOEPI BASES REVISION 12/10/14LOEP2 BASES REVISION 07/23/12LOEP3 BASES REVISION 12/10/14LOEP4 BASES REVISION 12/10/14LOEP5 BASES REVISION 12/10/14LOEP6 BASES REVISION 05/16/14LOEP7 BASES REVISION 12/10/14LOEP8 BASES REVISION 01/14/15LOEP9 BASES REVISION 06/13/14LOEP 10 BASES REVISION 05/16/12LOEP 11 BASES REVISION 12/10/14LOEP12 BASES REVISION 12/10/14LOEP 13 BASES REVISION 09/03/14LOEP14 BASES REVISION 08/28/14LOEP 15 BASES REVISION 05/16/12LOEP16 BASES REVISION 11/05/14LOEP17 BASES REVISION 11/05/14i BASES REVISION 06/03/11ii 363/365/364 10/29/08iii 355/357/356 04/02/07iv BASES REVISION 09/03/14B 2.1.1-1 BASES REVISION 05/31/12B 2.1.1-2 BASES REVISION 05/31/12B 2.1.1-3 BASES REVISION 05/31/12B 2.1.1-4 BASES REVISION 05/31/12B 2.1.2-1 BASES REVISION 02/06/14B 2.1.2-2 BASES REVISION 02/06/14B 2.1.2-3 BASES REVISION 02/06/14B 3.0-1 356/358/357 04/02/07B 3.0-2 BASES REVISION 10/23/03B 3.0-3 BASES REVISION 10/23/03B 3.0-4 BASES REVISION 10/23/03B 3.0-5 BASES REVISION 10/23/03B 3.0-6 BASES REVISION 10/23/03B 3.0-7 BASES REVISION 10/23/03B 3.0-8 BASES REVISION 10/23/03LOEPI OCONEE NUCLEAR STATIONTECHNICAL SPECIFICATIONS-BASES REVISED 01/14/15LIST OF EFFECTIVE PAGESPAGE AMENDMENT BASES REVISION DATEB 3.0-9 356/358/357 04/02/07B 3.0-10 356/358/357 04/02/07B 3.0-11 356/358/357 04/02/07B 3.0-12 356/358/357 04/02/07B 3.0-13 BASES REVISION 10/20/11B 3.0-14 BASES REVISION 10/23/03B 3.0-15 BASES REVISION 10/23/03B 3.1.1-1 BASES REVISION 05/16/12B 3.1.1-2 BASES REVISION 05/16/12B 3.1.1-3 BASES REVISION 05/16/12B3.1.1-4 BASES REVISION 05/16/12B 3.1.2-1 BASES REVISION 05/16/12B 3.1.2-2 BASES REVISION 05/16/12B 3.1.2-3 BASES REVISION 05/16/12B 3.1.2-4 BASES REVISION 05/16/12B 3.1.2-5 BASES REVISION 05/16/12B 3.1.3-1 BASES REVISION 06/02/99B 3.1.3-2 BASES REVISION 03/27/99B 3.1.3-3 300/300/300 12/16/98B 3.1.3-4 300/300/300 12/16/98B 3.1.4-1 BASES REVISION 07/23/12B 3.1.4-2 BASES REVISION 07/23/12B 3.1.4-3 BASES REVISION 07/23/12B 3.1.4-4 BASES REVISION 07/23/12B 3.1.4-5 BASES REVISION 07/23/12B 3.1.4-6 BASES REVISION 07/23/12B 3.1.4-7 BASES REVISION 07/23/12B 3.1.4-8 BASES REVISION 07/23/12B 3.1.4-9 BASES REVISION 07/23/12B 3.1.5-1 BASES REVISION 05/16/12B 3.1.5-2 BASES REVISION 05/16/12B 3.1.5-3 BASES REVISION 05/16/12B 3.1.5-4 BASES REVISION 05/16/12B 3.1.6-1 BASES REVISION 07/23/12B 3.1.6-2 BASES REVISION 07/23/12B 3.1.6-3 BASES REVISION 07/23/12B 3.1.6-4 DELETE BASES REV. 07/23/12LOEP2 OCONEE NUCLEAR STATIONTECHNICAL SPECIFICATIONS-BASES REVISED 01/14/15LIST OF EFFECTIVE PAGESPAGE AMENDMENT BASES REVISION DATEB 3.1.7-1 BASES REVISION 07/23/12B 3.1.7-2 BASES REVISION 07/23/12B 3.1.7-3 BASES REVISION 07/23/12B 3.1.7-4 BASES REVISION 07/23/12B 3.1.8-1 BASES REVISION 05/16/12B 3.1.8-2 BASES REVISION 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3.3.1-5 BASES REVISION 12/10/14B 3.3.1-6 BASES REVISION 12/10/14LOEP3 OCONEE NUCLEAR STATIONTECHNICAL SPECIFICATIONS-BASES REVISED 01/14/15LIST OF EFFECTIVE PAGESPAGEB 3.3.1-7B 3.3.1-8B 3.3.1-9B 3.3.1-10B 3.3.1-11B 3.3.1-12B 3.3.1-13B 3.3.1-14B 3.3.1-15B 3.3.1-16B 3.3.1-17B 3.3.1-18B 3.3.1-19B 3.3.1-20B 3.3.1-21B 3.3.1-22B 3.3.1-23B 3.3.1-24B 3.3.1-25B.3.3.1-26 B.3.3.1-27 B.3.3.1-28 B.3.3.1-29 B.3.3.1-30 B 3.3.2-1B 3.3.2-2B 3.3.2-3B 3.3.3-1B 3.3.3-2B 3.3.3-3B 3.3.3-4B.3.3.3-5 B.3.3.3-6 B 3.3.4-1B 3.3.4-2B 3.3.4-3B 3.3.4-4B 3.3.5-1B 3.3.5-2B 3.3.5-3B 3.3.5-4AMENDMENT BASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONDELETEDDELETEDDELETEDDELETEDBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONDELETEDDELETEDBASES 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REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISION DATE05/16/1205/16/1205/16/1212/18/0712/18/0712/18/0712/18/0705/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/1205/16/12LOEP 15 OCONEE NUCLEAR STATIONTECHNICAL SPECIFICATIONS-BASES REVISED 01/14/15LIST OF EFFECTIVE PAGESPAGEB 3.9.2-1B 3.9.2-2B 3.9.2-3B 3.9.2-4B 3.9.3-1B 3.9.3-2B 3.9.3-3B 3.9.3-4B 3.9.3-5B 3.9.4-1B 3.9.4-2B 3.9.4-3B 3.9.4-4B 3.9.5-1B 3.9.5-2B 3.9.5-3B 3.9.5-4B 3.9.6-1B 3.9.6-2B 3.9.6-3B 3.9.7-1B 3.9.7-2B 3.9.7-3B 3.9.8-1B 3.9.8-2B 3.9.8-3B 3.10.1-1B 3.10.1-2B 3.10.1-3B 3.10.1-4B 3.10.1-5B 3.10.1-6B 3.10.1-7B 3.10.1-8B 3.10.1-9B 3.10.1-10 B 3.10.1-11 B 3.10.1-12 B 3.10.1-13 B 3.10.1-14 B 3.10.1-15 B 3.10.1-16 B 3.10.1-17 AMENDMENT BASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISIONBASES REVISION 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These loads areintroduced by startup (heatup) and shutdown (cooldown) operations, powertransients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
Figures 3.4.3-1 through 3.4.3-9 contain P/T limit curves for heatup,cooldown, and leak and hydrostatic (LH) testing.
Tables 3.4.3-1 and3.4.3 2 contain data for the maximum rate of change of reactor coolanttemperature.
The minimum temperature indicated in the P/T limit curvesand tables of 60OF is the lowest unirradiated nil ductility reference temperature (RTNDT) of all materials in the reactor vessel. Thistemperature (600F) is the minimum allowable reactor pressure vesseltemperature if any head closure stud is not fully detensioned.
There is nominimum allowable temperature limit for the reactor vessel if all of thestuds are fully detensioned.
Figures 3.4.3-1, 3.4.3-2, 3.4.3-4, 3.4.3-5, 3.4.3-7 and 3.4.3-8 define anacceptable region for normal operation.
The usual use of the curves isoperational guidance during heatup or cooldown maneuvering, whenpressure and temperature indications are monitored and compared to theapplicable curve to determine that operation is within the allowable region.The LCO establishes operating limits that provide a margin to brittle failureof the reactor vessel and piping of the reactor coolant pressure boundary(RCPB). The vessel is the component most subject to brittle failure, andthe LCO limits apply mainly to the vessel. The limits do not apply to thepressurizer, which has different design characteristics and operating functions.
10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limitsfor material fracture toughness requirements of the RCPB materials.
Reference 1 requires an adequate margin to brittle failure during normaloperation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code, Section III,Appendix G (Ref. 2).Linear elastic fracture mechanics (LEFM) methodology is used todetermine the stresses and material toughness at locations within theRCPB. The LEFM methodology follows the guidance given by 10 CFR 50,Appendix G; ASME Code, Section III, Appendix G; and Regulatory Guide 1.99 (Ref. 3).OCONEE UNITS 1, 2, & 3B 3.4.3-1BASES REVISION DATED 01/14/15 1
RCS P/T LimitsB 3.4.3BASESBACKGROUND Material toughness properties of the ferritic materials of the reactor(continued) vessel are determined in accordance with ASTM E 185 (Ref. 4), andadditional reactor vessel requirements.
These properties are thenevaluated in accordance with Reference 2.The actual shift in the nil ductility reference temperature (RTNDT) of thevessel material will be established periodically by evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185(Ref. 5) and Appendix H of 10 CFR 50 (Ref. 5). The operating P/T limitcurves will be adjusted, as necessary, based on the evaluation findingsand the recommendations of Reference 2.The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vesseland head that are the most restrictive.
At any specific
: pressure, temperature, and temperature rate of change, one location within thereactor vessel will dictate the most restrictive limit. Across the span of theP/T limit curves, different locations are more restrictive, and, thus, thecurves are composites of the most restrictive regions.The heatup curve represents a different set of restrictions than thecooldown curve because the directions of the thermal gradients throughthe vessel wall are reversed.
The thermal gradient reversal alters thelocation of the tensile stress between the outer and inner walls.The calculation to generate the LH testing curve uses different safetyfactors (per Ref. 2) than the heatup and cooldown curves.The P/T limit curves and associated temperature rate of change limits aredeveloped in conjunction with stress analyses for large numbers ofoperating cycles and provide conservative margins to nonductile failure.Although created to provide limits for these specific normal operations, thecurves also can be used to determine if an evaluation is necessary for anabnormal transient.
As stated in the tables associated with this LCO, reactor coolant (RC)temperature is cold leg temperature if one or more RC pumps are inoperation; otherwise, it is the LPI cooler outlet temperature.
An analysisexamined the effects of initiating flow through a previously idle LPI train(i.e. either placing a train of LPI in operation or swapping from one train tothe other) when none of the RC pumps are operating.
The analysisassumed the initial temperature of the fluid entering the vessel to be thelowest expected temperature in an idle LPI cooler. As RC fluid is pumpedthrough the system and returns to the reactor vessel, the temperature increases to a "stable" value. The duration of the temperature excursion is dependent on LPI flow and volume of the piping system. This analysishas determined that the brief temperature excursion caused by the fluidinitially in the idle LPI train can be accommodated if, at the time the LPIheader is put in service, the RCS pressure is less than 295 psig(Instrument Uncertainty Adjusted).
This value is less limiting than theOCONEE UNITS 1, 2, & 3B 3.4.3-2BASES REVISION DATED 01/14/15 I
RCS P/T LimitsB 3.4.3BAS ESBACKGROUND (continued)
LPI initiation pressure limit imposed by procedures to protect the LPIsystem from overpressure.
The brief temperature excursion does notplace the reactor vessel outside of the bounds of the stress analyses.
The criticality limit curve includes the Reference 1 requirement that it be40°F above the heatup curve or the cooldown curve, and not less than theminimum permissible temperature for LH testing.
: However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO3.4.2, "RCS Minimum Temperature for Criticality."
The consequence of violating the LCO limits is that the RCS has beenoperated under conditions that can result in brittle failure of the RCPB,possibly leading to a nonisolable leak or loss of coolant accident.
In theevent these limits are exceeded, an evaluation must be performed todetermine the effect on the structural integrity of the RCPB components.
The ASME Code, Section XI, Appendix E (Ref. 6) provides arecommended methodology for evaluating an operating event that causesan excursion outside the limits.APPLICABLE SAFETY ANALYSESThe P/T limits are not derived from accident analyses.
They areprescribed during normal operation to avoid encountering
: pressure, temperature, and temperature rate of change conditions that might causeundetected flaws to propagate and cause nonductile failure of the RCPB,an unanalyzed condition.
Reference 1 establishes the methodology fordetermining the P/T limits. Since the P/T limits are not derived from anyaccident
: analysis, there are no acceptance limits related to the P/T limits.Rather, the P/T limits are acceptance limits themselves since theypreclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36 (Ref. 7).LCOThe three elements of this LCO are:a. The limit curves for heatup and cooldown,
: b. Limits on the rate of change of temperature, andc. Allowable RC pump combinations.
The LCO is modified by three Notes. Note 1 states that for leak tests ofthe RCS and leak tests of connected systems where RCS pressure andtemperature are controlling, the RCS may be pressurized to the limits ofthe specified figures.
Note 2 states that for thermal steady state hydrotests required by ASME Section XI RCS may be pressurized to the limitsSpecification 2.1.2 and the specified figures.
The limits on the rate ofchange of reactor coolant temperature RCS P/T Limits are the same onesOCONEE UNITS 1, 2, & 3B 3.4.3-3BASES REVISION DATED 01/14/15 I
RCS P/T LimitsB 3.4.3BASESLCO used for normal heatup and cooldown operations.
Note 3 states the RCS(continued)
P/T limits are not applicable to the pressurizer.
The LCO limits apply to all components of the RCS, except thepressurizer.
These limits define allowable operating regions and permit alarge number of operating cycles while providing a wide margin tononductile failure.Table 3.4.3-1 includes temperature rate of change limits with allowable pump combinations for RCS heatup while Table 3.4.3-2 includestemperature rate of change limits with allowable pump combinations forRCS cooldown.
The breakpoints between temperature rate of changelimits in these two tables are selected to limit reactor vessel thermalgradients to acceptable limits. The breakpoint between allowable pumpcombinations was selected based on operational requirements and areused to determine the change of RCS pressure associated with thechange in number of operating reactor coolant pumps.The limits for the rate of change of temperature control the thermalgradient through the vessel wall and are used as inputs for calculating theheatup, cooldown, and LH P/T limit curves. Thus, the LCO for the rate ofchange of temperature restricts stresses caused by thermal gradients andalso ensures the validity of the P/T limit curves.The limits on allowable RC pump combinations controls the pressuredifferential between the vessel wall and the pressure measurement pointand are used as inputs for calculating the heatup, cooldown and LH P/Tlimit curves. Thus, the LCO for the allowable RC pump combinations restricts the pressure at the vessel wall and ensures the validity of the P/Tlimit curves.Heatup and Cooldown Rate limits are specified in TS Table 3.4.3-1"Operational Requirements for Unit Heatup" and TS Table 3.4.3-2"Operational Requirements for Unit Cooldown."
These limits are specified as a change in temperature for "any" time period. As such, the Heatup orCooldown period is a rolling period and is required to be considered atany point in time, i.e., the beginning, middle, or end of the period underevaluation.
This action is required to ensure the heatup or cooldown ratelimit meets design limits.The LPI cooler outlet temperature during the brief period of stabilization does not need to be considered when determining heatup or cooldownrates or RCS P/T conditions when an LPI train is placed in operation withno operating RCPs. The period of stabilization is the time required to fullydisplace the stagnant fluid in the idle LPI train. The time required for LCOstabilization is a function of LPI flow rate. Operating procedures controlboth placing a train of LPI in service and swapping trains of LPI to limit theduration of the temperature transient to a value that has been shown to beacceptable.
OCONEE UNITS 1, 2, & 3B 3.4.3-4BASES REVISION DATED 01/14/15 I
RCS P/T LimitsB 3.4.3BASESLCO(continued)
Violating the LCO limits places the reactor vessel outside of the bounds ofthe stress analyses and can increase stresses in other RCPBcomponents.
The consequences depend on several factors, as follows:a. The severity of the departure from the allowable operating P/Tregime or the severity of the rate of change of temperature;
: b. The length of time the limits were violated (longer violations allowthe temperature gradient in the thick vessel walls to become morepronounced);
andc. The existences, sizes, and orientations of flaws in the vesselmaterial.
APPLICABILITY The RCS P/T limits Specification provides a definition of acceptable operation for prevention of nonductile failure in accordance with10 CFR 50, Appendix G (Ref. 1). Although the P/T limits were developed to provide guidance for operation during heatup or cooldown (MODES 3,4, and 5) or LH testing, their applicability is at all times in keeping with theconcern for nonductile failure.
The limits do not apply to the pressurizer.
During MODES 1 and 2, other Technical Specifications provide limits foroperation that can be more restrictive than or can supplement these P/Tlimits. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure fromNucleate Boiling (DNB) Limits";
LCO 3.4.2, "RCS Minimum Temperature for Criticality";
and Safety Limit (SL) 2.1, "SLs," also provide operational restrictions for pressure and temperature and maximum pressure.
MODES 1 and 2 are above the temperature range of concern fornonductile
: failure, and stress analyses have been performed for normalmaneuvering
: profiles, such as power ascension or descent.ACTIONSA.1 and A.2Operation outside the P/T limits during MODE 1, 2, 3, or 4 must becorrected so that the RCPB is returned to a condition that has beenverified by stress analyses.
The 30 minute Completion Time reflects the urgency of restoring theparameters to within the analyzed range. Most violations will not besevere, and the activity can be accomplished in this time in a controlled manner.Besides restoring operation to within limits, an evaluation is required todetermine if RCS operation can continue.
The evaluation must verify theRCPB integrity remains acceptable and must be completed beforecontinuing operation.
Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, newanalyses, or inspection of the components.
The evaluation must beOCONEE UNITS 1, 2, & 3B 3.4.3-5BASES REVISION DATED 01/14/15 I
RCS P/T LimitsB 3.4.3BASESACTIONS A.1 and A.2 (continued) completed, documented, and approved in accordance with established plant procedures and administrative controls.
ASME Code, Section XI, Appendix E (Ref. 6) may be used to support theevaluation.
: However, its use is restricted to evaluation of the vesselbeltline.
The evaluation must extend to all components of the RCPB.The 72 hour Completion Time is reasonable to accomplish the evaluation.
The evaluation for a mild violation is possible within this time, but moresevere violations may require special, event specific stress analyses orinspections.
A favorable evaluation must be completed before continuing to operate.Condition A is modified by a Note requiring Required Action A.2 to becompleted whenever the Condition is entered.
The Note emphasizes theneed to perform the evaluation of the effects of the excursion outside theallowable limits. Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may haveaffected the RCPB integrity.
B.1 and B.2If a Required Action and associated Completion Time of Condition A arenot met, the unit must be brought to a lower MODE because:
(a) the RCSremained in an unacceptable pressure and temperature region for anextended period of increased stress, or (b) a sufficiently severe eventcaused entry into an unacceptable region. Either possibility indicates aneed for more careful examination of the event, best accomplished withthe RCS at reduced pressure and temperature.
With reduced pressureand temperature conditions, the possibility of propagation of undetected flaws is decreased.
If the required restoration activity cannot be accomplished within30 minutes, Required Action B.1 and Required Action B.2 must beimplemented to reduce pressure and temperature.
If the required evaluation for continued operation cannot be accomplished within 72 hours, or the results are indeterminate or unfavorable, actionmust proceed to reduce pressure and temperature as specified inRequired Actions B.1 and B.2. A favorable evaluation must be completed and documented before returning to operating pressure and temperature conditions.
: However, if the favorable evaluation is accomplished whilereducing pressure and temperature conditions, a return to poweroperation may be considered without completing Required Action B.2.Pressure and temperature are reduced by bringing the unit to MODE 3within 12 hours and to MODE 5 within 36 hours. The allowed Completion OCONEE UNITS 1, 2, & 3B 3.4.3-6BASES REVISION DATED 01/14/15 I
RCS P/T LimitsB 3.4.3BASESACTIONS B.1 and B.2 (continued)
Times are reasonable, based on operating experience, to reach therequired MODE from full power conditions in an orderly manner andwithout challenging unit systems.C.1 and C.2Actions must be initiated immediately to correct operation outside of theP/T limits at times other than MODE 1, 2, 3, or 4, so that the RCPB isreturned to a condition that has been verified acceptable by stressanalysis.
The immediate Completion Time reflects the urgency of initiating action torestore the parameters to within the analyzed range. Most violations willnot be severe, and the activity can be accomplished within this time in acontrolled manner.In addition to restoring operation to within limits, an evaluation is requiredto determine if RCS operation can continue.
The evaluation must verifythat the RCPB integrity remains acceptable and must be completed priorto entry into MODE 4. Several methods may be used, including comparison with pre-analyzed transients in the stress analysis, orinspection of the components.
ASME Code, Section Xl, Appendix E (Ref. 6), may also be used tosupport the evaluation.
: However, its use is restricted to evaluation of thevessel beltline.
Condition C is modified by a Note requiring Required Action C.2 to becompleted whenever the Condition is entered.
The Note emphasizes theneed to perform the evaluation of the effects of the excursion outside theallowable limits. Restoration alone, per Required Action C.1, isinsufficient because higher than analyzed stresses may have occurredand may have affected RCPB integrity.
SURVEILLANCE SR 3.4.3.1REQUIREMENTS Verification that operation is within limits is required when RCS pressureor temperature conditions are undergoing planned changes.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.Surveillance for heatup, cooldown, or LH testing may be discontinued when the definition given in the relevant plant procedure for ending theactivity is satisfied.
OCONEE UNITS 1, 2, & 3B 3.4.3-7BASES REVISION DATED 01/14/15 I
RCS P/T LimitsB 3.4.3BASESSURVEILLANCE REQUIREMENTS SR 3.4.3.1 (continued)
This SR is modified by a Note that requires this SR to be performed onlyduring system heatup, cooldown, and LH testing.REFERENCES 1.2.3.4.5.6.7.10 CFR 50, Appendix G.ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.Regulatory Guide 1.99, Revision 2, May 1988.ASTM E 185-82, July 1982.10 CFR 50, Appendix H.ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.10 CFR 50.36.OCONEE UNITS 1, 2, & 3B 3.4.3-8BASES REVISION DATED 01/14/15 I
RPS Instrumentation B 3.3.1B 3.3 INSTRUMENTATION B 3.3.1 Reactor Protective System (RPS) Instrumentation BASESBACKGROUND The RPS initiates a reactor trip to protect against violating the core fueldesign limits and the Reactor Coolant System (RCS) pressure boundaryduring anticipated transients.
By tripping the reactor, the RPS also assiststhe Engineered Safeguards (ES) Systems in mitigating accidents.
The protective and monitoring systems have been designed to assure safeoperation of the reactor.
This is achieved by specifying limiting safetysystem settings (LSSS) in terms of parameters directly monitored by theRPS, as well as the LCOs on other reactor system parameters andequipment performance.
The LSSS, defined in this Specification as the Allowable Value, inconjunction with the LCOs, establishes the threshold for protective systemaction to prevent exceeding acceptable limits during accidents ortransients.
During anticipated transients, which are those events expected to occurone or more times during the unit's life, the acceptable limit is:a. The departure from nucleate boiling ratio (DNBR) shall bemaintained above the Safety Limit (SL) value;b. Fuel centerline melt shall not occur; andc. The RCS pressure SL of 2750 psia shall not be exceeded.
Maintaining the parameters within the above values ensures that the offsitedose will be within the 10 CFR 20 and 10 CFR 100 criteria duringanticipated transients.
Accidents are events that are analyzed even thoughthey are not expected to occur during the unit's life. The acceptable limitduring accidents is that the offsite dose shall be maintained withinreference 10 CFR 100 limits. Meeting the acceptable dose limit for anaccident category is considered having acceptable consequences for thatevent.OCONEE UNITS 1, 2, & 3B 3.3.1 -112/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND RPS Overview(continued)
The RPS consists of four separate redundant protective channels thatreceive inputs of neutron flux, RCS pressure, RCS flow, RCS temperature, RCS pump status, reactor building (RB) pressure, main feedwater (MFW)pump turbines status, and main turbine status.Figure 7.1 of UFSAR, Chapter 7 (Ref. 1), shows the arrangement of atypical RPS protective channel.
A protective channel is composed ofmeasurement
: channels, a manual trip channel, a reactor trip component (RTC), and a control rod drive (CRD) trip device. LCO 3.3.1 providesrequirements for the individual measurement channels.
These channelsencompass all equipment and electronics from the point at which themeasured parameter is sensed through the processor output trip devicesin the trip string. LCO 3.3.2, "Reactor Protective System (RPS) ManualReactor Trip," LCO 3.3.3, "Reactor Protective System (RPS) -Reactor TripComponent (RTC)," and LCO 3.3.4, "Control Rod Drive (CRD) TripDevices,"
discuss the remaining RPS elements.
The RPS instrumentation measures critical unit parameters and comparesthese to predetermined setpoints.
If the setpoint for a parameter input to a single channel (for example, theRC high pressure input to Channel A) is exceeded, a channel trip doesnot occur. Due to the inter-channel communication, all 4 RPS channelsrecognize that this parameter input has been exceeded for one channel.However, due to the 2.MIN/2.MAX logic within the system, the sameparameter input setpoint for one of the other three channels must beexceeded before channel trips occur. Again, due to the inter-channel communication, all 4 RPS channels will then trip since the 2.MIN/2.MAX condition has been satisfied.
The RTS consists of four AC Trip Breakers arranged in two parallelcombinations of two breakers each. Each path provides independent power to the CRD motors. Either path can provide sufficient power tooperate all CRD's. Two separate power paths to the CRD's ensure that asingle failure that opens one path will not cause an unwanted reactor trip.The RPS consists of four independent protective channels (A, B, C, and D).Each RPS protective channel contains the sensor input modules, aprotective channel computer, output modules, four hardwired (energized during power operations) reactor trip relays (RTRs) (A, B, C, and D) andtheir associated 120 VAC contacts (closed when RTR is energized).
OCONEE UNITS 1, 2, & 3B 3.3.1-212/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND RPS Overview (continued)
Protective channel A controls the channel A RTR and also controls the ARTR in channels B, C, and D. Likewise, channels B, C and D control therespective RTR in each of the four channels.
Each energized RTR (A, B,C, and D) in each RPS channel A, B, C, and D maintains two closed 120VAC contacts.
One contact from each RTR is configured in two separateredundant output trip actuation logic schemes.
Each output trip actuation logic scheme contains a contact from each of the four RTRs in the fourchannels.
This configuration results in a two-out-of-four coincidence reactor trip logic. If any channel protective set initiates a trip signal, therespective four RTRs (one in each of the four channels) de-energize andopen the respective contacts.
The outputs from the RTR contacts interrupt the 120 VAC power to the CRD trip devices.Three of the four RPS protective channel computers (A, B, and C) alsoperform a redundant Engineered Safeguards Protective System (ESPS)logic function.
Therefore, three of the four RPS protective channelscalculate both RPS and ESPS functions, and the fourth RPS channel Dcalculates only RPS functions.
See Technical Specification Bases sectionB 3.3.5 for additional discussion of the ESPS protective channels and theduplicated ESPS functions performed by the RPS protective channels.
The reactor is tripped by opening the reactor trip breakers.
There are three bypasses:
shutdown bypass, manual bypass, and channel Itrip function bypass. The shutdown bypass and the manual bypass areinitiated by use of a keyswitch located in the respective RPS channelcabinet.
The Shutdown bypass allows the withdrawal of safety rods forSDM availability and rapid negative reactivity insertion during unitcooldowns or heatups.
The manual bypass allows putting a complete RPSchannel into bypass for maintenance activities.
This includes the plannedpower-down of the bypassed RPS channel computer.
If the complete RPSchannel is powered down, the manual bypass condition cannot bemaintained.
That RPS channel output signal goes to "trip" and the manualbypass Unit Statalarm window will not illuminate.
The channel trip functionbypass allows an individual channel trip function in any RPS channel to bebypassed through the use of the RPS screens of the Graphical ServiceMonitor (GSM). The GSM is located on the Service Unit.The RPS operates from the instrumentation channels discussed next. Thespecific relationship between measurement channels and protective channels differs from parameter to parameter.
Three basic configurations are used:OCONEE UNITS 1, 2, & 3B 3.3.1-312/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND RPS Overview (continued)
: a. Four completely redundant measurements (e.g., reactor coolantflow) with one channel input to each protective channel;b. Four channels that provide similar, but not identical, measurements (e.g., power range nuclear instrumentation where each RPSchannel monitors a different quadrant),
with one channel input toeach protective channel; andc. Redundant measurements with combinational trip logic inside theprotective channels and the combined output provided to eachprotective channel (e.g., main feedwater pump turbines tripinstrumentation).
These arrangements and the relationship of instrumentation channels totrip Functions are discussed next to assist in understanding the overalleffect of instrumentation channel failure.Power Range Nuclear Instrumentation Power Range Nuclear Instrumentation channels provide inputs to thefollowing trip Functions:
: 1. Nuclear Overpower
: a. Nuclear Overpower
-High Setpoint;
: b. Nuclear Overpower
-Low Setpoint;
: 7. Reactor Coolant Pump to Power;8. Nuclear Overpower Flux/Flow Imbalance;
: 9. Main Turbine Trip (Hydraulic Fluid Pressure);
and10. Loss of Main Feedwater (LOMFW) Pump Turbines (Hydraulic OilPressure).
OCONEE UNITS 1, 2, & 3B 3.3.1-412/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND Power Range Nuclear Instrumentation (continued)
The power range instrumentation has four linear level channels, one foreach core quadrant.
Each channel feeds one RPS protective channel.Each channel originates in a detector assembly containing twouncompensated ion chambers.
The ion chambers are positioned torepresent the top half and bottom half of the core. The individual currentsfrom the chambers are fed to individual linear amplifiers.
The summation of the top and bottom is the total reactor power. The difference of the topminus the bottom neutron signal is the measured AXIAL POWERIMBALANCE for the associated core quadrant.
Reactor Coolant System Outlet Temperature The Reactor Coolant System Outlet Temperature provides input to thefollowing Functions:
: 2. RCS High Outlet Temperature; and5. RCS Variable Low Pressure.
The RCS Outlet Temperature is measured by two resistance temperature detection elements in each hot leg, for a total of four. One temperature detection element is associated with each protective channel.Reactor Coolant System PressureThe Reactor Coolant System Pressure provides input to the following Functions:
: 3. RCS High Pressure;
: 4. RCS Low Pressure;
: 5. RCS Variable Low Pressure; and11. Shutdown Bypass RCS High Pressure.
The RPS inputs of reactor coolant pressure are provided by two pressuretransmitters in each hot leg, for a total of four. One sensor is associated with each protective channel.OCONEE UNITS 1, 2, & 3B 3.3.1-512/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND Reactor Building Pressure(continued)
The Reactor Building Pressure measurements provide input only to theReactor Building High Pressure trip, Function
: 6. There are four RB HighPressure
: sensors, one associated with each protective channel.Reactor Coolant Pump Power Monitoring Reactor coolant pump power monitors are inputs to the Reactor CoolantPump to Power trip, Function
: 7. Each RCP has a RCP Power Monitor(RCPPM),
which monitors the electrical power and breaker status of eachpump motor to determine if it is running.
Each RCPPM provides inputs toall four RPS channels.
Reactor Coolant System FlowThe Reactor Coolant System Flow measurements are an input to theNuclear Overpower Flux/Flow Imbalance trip, Function
: 8. The reactorcoolant flow inputs to the RPS are provided by eight high accuracydifferential pressure transmitters, four on each loop, which measure flowthrough calibrated flow tubes. One flow input in each loop is associated with each protective channel.Main Turbine Hydraulic Fluid PressureMain Turbine Hydraulic Fluid Pressure is an input to the Main Turbine Trip(Hydraulic Fluid Pressure) reactor trip, Function
: 9. Each of the fourprotective channels receives turbine status information from one of the fourpressure switches monitoring main turbine hydraulic fluid pressure.
Eachprotective channel continuously monitors the status of the contact inputsand initiates an RPS trip when a main turbine trip is indicated.
Feedwater Pump Turbine Hydraulic Oil PressureFeedwater Pump Turbine Hydraulic Oil Pressure is an input to the Loss ofMain Feedwater Pumps (Hydraulic Oil Pressure) trip, Function 10.Hydraulic Oil pressure is measured by four switches on each feedwater pump turbine.
One switch on each pump turbine is associated with eachprotective channel.OCONEE UNITS 1, 2, & 3B 3.3.1-612/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND Feedwater Pump Turbine Hydraulic Oil Pressure (continued)
Each RPS channel receives a contact input from both Feedwater PumpTurbines (A and B) Hydraulic Oil Pressure switches.
When the switchesfrom both turbines indicate that the associated Turbine Hydraulic OilPressure is low (turbine has tripped),
a reactor trip signal is initiated onthat channel.RPS BypassesThe RPS is designed with three types of bypasses:
shutdown bypass,manual bypass and channel trip function bypass.Each bypass is discussed next.Shutdown BypassDuring unit cooldown and heatup, it is desirable to leave the safety rods atleast partially withdrawn to provide shutdown capabilities in the event ofunusual positive reactivity additions (moderator
: dilution, etc.).However, the unit is also depressurized as coolant temperature isdecreased.
If the safety rods are withdrawn and coolant pressure isdecreased, an RCS Low Pressure trip will occur at 1800 psig and the rodswill fall into the core. To avoid this, the protective system allows theoperator to bypass the low pressure trip and maintain shutdowncapabilities.
During the cooldown and depressurization, the safety rods areinserted prior to the low pressure trip of 1800 psig. The RCS pressure isdecreased to less than 1720 psig, then each RPS channel is placed inshutdown bypass.A shutdown bypass signal is provided by the operator from the shutdownbypass keyswitch (status shall be indicated by a light). This actionbypasses the RCS Low Pressure trip, Nuclear Overpower Flux/Flow Imbalance trip, Reactor Coolant Pump to Power trip, and the RCS VariableLow Pressure trip, and inserts a new RCS High Pressure, 1720 psig trip.The operator can now withdraw the safety rods for additional rapidlyinsertable negative reactivity.
The insertion of the new high pressure trip performs two functions.
First,with a trip setpoint of 1720 psig, the processor output trip device preventsoperation at normal system pressure, 2155 psig, with a portion of the RPSbypassed.
The second function is to ensure that the bypass is removedprior to normal operation.
When the RCS pressure is increased during aOCONEE UNITS 1, 2, & 3B 3.3.1-712/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND Shutdown Bypass (continued) unit heatup, the safety rods are inserted prior to reaching 1720 psig. Theshutdown bypass is removed, which returns the RPS to normal, andsystem pressure is increased to greater than 1800 psig. The safety rodsare then withdrawn and remain at the full out condition for the rest of theheatup.In addition to the Shutdown Bypass RCS High Pressure trip, the High FluxReactor Trip setpoint is automatically lowered to less than 5% when theoperator closes the shutdown bypass keyswitch.
This provides a backupto the Shutdown Bypass RCS High Pressure trip and allows testing whilepreventing the generation of any significant amount of power.Manual BypassThe RPS Manual Bypass allows putting the complete RPS channel intobypass for maintenance activities.
Placing the RPS channel in bypass doesnot power-down the computer.
If it is necessary to power-down thecomputer for one channel, the Manual Bypass keyswitch is used to keepthe four RTRs associated with the respective channel energized while thechannel computer is powered down. To place a protective channel inmanual bypass, the other three channels must not be in manual bypass orotherwise inoperable (e.g., a channel trip function in bypass).The RPS Manual Bypass status information is sent to the Unit Statalarm panel (hardwired output of the RPS Channel computer and in parallel as ahardwired signal from a keyswitch contact in case the computer is powereddown) and is sent to the plant Operator Aid Computer (OAC) via agateway.If the complete RPS cabinet is powered down, the Manual Bypasscondition cannot be maintained.
That RPS channel output signal goes to"trip" and the Manual Bypass Unit Statalarm window will not illuminate.
Channel Trip Function BypassAn individual Channel Trip Function Bypass allows placing one trip functionin bypass for maintenance activities through the RPS GSM screens.
Thisallows the remaining trip functions in the channel to remain operable whilethe channel input device for the affected channel is inoperable.
Operation to put functions in bypass is administratively controlled sincethere is no interlock to prevent placing functions in multiple channels inbypass. Channel trip functions may be placed in bypass in only one RPSchannel at a time.OCONEE UNITS 1, 2, & 3B 3.3.1-812/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND Parameter Change Enable Mode(continued)
Parameter Change Enable Mode allows each RPS instrument inputchannel processor to be placed in different operating modes through theuse of the Parameter Change Enable keyswitches and commands from theService Unit. Each protective channel has a keyswitch located in thatchannel's cabinet pair.Placing RPS Channels A, B, or C in Parameter Change Enable Modethrough the use of the "Parameter Change Enable" keyswitch will alsoplace the corresponding ESPS Channels Al, B1 or Cl in Parameter Change Enable Mode.When a keyswitch is placed from the normal Operating Mode position tothe Parameter Change Enable Mode position:
* The processors continue with normal operation.
* A permissive is provided that allows the Service Unit to be used tochange the operating mode of the processors associated with thatkeyswitch.
With the keyswitch in the Parameter Change Enable Position thefollowing modes of operation are allowed for processors:
* Normal Operation
-with permissive for operating mode change.* Parameterization
-allows changes to specific parameters (exampleplacing a parameter into a tripped condition or performing ReactorTrip Relay testing).
* Function Test -for disabling the application function and forcingoutput signal for testing purposes (normally not used).* Diagnostics
-for downloading new application software.
The Function Test and Diagnostics modes result in the processor ceasingits cyclic processing of the application functions.
Entry into these modesfirst requires entry into Parameterization mode and setting a separateparameter.
When a keyswitch is placed in the Parameter Change Enable ModePosition for any activity, the affected processor shall first be declared outof service.
In addition to declaring the processor out of service (1) theaffected RPS channel shall be bypassed and (2) either the affectedESPS input channel (Al, B1, or Cl) shall be tripped OR the ESPS Set 1voters shall be placed in Bypass for the following activities:
* Loading or revising the software in a processor.
" Changing parameters via the RPS High Flux Trip (Variable Setpoint) screen at the Service Unit.OCONEE UNITS 1, 2, & 3B 3.3.1-912/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND Parameter Change Enable Mode (continued)
Changing parameters via the RPS Flux/Flow/Imbalance Parameters screen at the Service Unit.Only one RPS channel at a time is allowed to be placed into Parameter Change Enable Mode Position for these activities.
Each Parameter Change Enable keyswitch status information is sent tothe Statalarm panel and to the OAC via the Gateway.RPS Parameter Change Enable keyswitches are administratively controlled (there are no hardware or software interlocks betweenchannels).
Trip Setpoints/Allowable ValueThe Allowable Value and trip setpoint are based on the analytical limitsstated in UFSAR, Chapter 15 (Ref. 2). The selection of the Allowable Value and associated trip setpoint is such that adequate protection isprovided when all sensor and processing time delays are taken intoaccount.
To allow for calibration tolerances, instrumentation uncertainties, instrument drift, and severe environment errors for those RPS channelsthat must function in harsh environments as defined by 10 CFR 50.49(Ref. 3), the Allowable Values specified in Table 3.3.1-1 in theaccompanying LCO are conservative with respect to the analytical limits toaccount for all known uncertainties for each channel.
The actual tripsetpoint entered into the processor output trip device is more conservative than that specified by the Allowable Value to account for changes inrandom measurement errors detectable by a CHANNEL CALIBRATION.
A channel is inoperable if its actual trip setpoint is not within its requiredAllowable Value. All field sensors and signal processing equipment forthese channels are assumed to operate within the allowances of theseuncertainty magnitudes.
The trip setpoints are the nominal values at whichthe processor output trip devices are set. Any processor output tripdevice is considered to be properly adjusted when the "as left" value iswithin the band for CHANNEL CALIBRATION accuracy.
A detaileddescription of the methodology used to determine the Allowable Value andassociated uncertainties is provided in Reference 4.Setpoints in conjunction with the Allowable Value ensure that the limits ofChapter 2.0, "Safety Limits,"
in the Technical Specifications are not violatedduring anticipated transients and that the consequences of accidents willbe acceptable, providing the unit is operated from within the LCOs at theonset of the anticipated transient or accident and the equipment functions OCONEE UNITS 1, 2, & 3B 3.3.1 -1012/10/14 1
RPS Instrumentation B 3.3.1BASESBACKGROUND Trip Setpoints/Allowable Value (continued) as designed.
Note that in LCO 3.3.1 the Allowable Values listed in Table3.3.1-1 for Functions 1 through 8 and 11 are the LSSS.With the exception of the RB High Pressure
: function, each channel istested online by manually retrieving the software setpoint to ensure it hasbeen entered correctly.
Signals into the system (from the field instrument or at the protective system cabinet) are applied during the channelcalibration to ensure that the instrumentation is within the specified allowance requirements.
APPLICABLE Each of the analyzed accidents and transients that require a reactor trip toSAFETY ANALYSES, meet the acceptance criteria can be detected by one or more RPSLCO, and Functions.
The accident analysis contained in the UFSAR, Chapter 15APPLICABILITY (Ref. 2), takes credit for most RPS trip Functions.
Functions notspecifically credited in the accident analysis were qualitatively credited inthe safety analysis and the NRC staff approved licensing basis for the unit.These Functions are high RB pressure, turbine trip, and loss of mainfeedwater.
These Functions may provide protection for conditions that donot require dynamic transient analysis to demonstrate Functionperformance.
These Functions also serve as backups to Functions thatwere credited in the safety analysis.
The LCO requires all instrumentation performing an RPS Function to beOPERABLE.
Failure of any instrument renders the affected channel(s) inoperable and reduces the reliability of the affected Functions.
The threechannels of each Function in Table 3.3.1-1 of the RPS instrumentation shall be OPERABLE during its specified Applicability to ensure that areactor trip will be actuated if needed. Additionally, during shutdownbypass with any CRD trip breaker closed, the applicable RPS Functions must also be available.
This ensures the capability to trip the withdrawn CONTROL RODS exists at all times that rod motion is possible.
The tripFunction channels specified in Table 3.3.1-1 are considered OPERABLEwhen all channel components necessary to provide a reactor trip arefunctional and in service for the required MODE or Other Specified Condition listed in Table 3.3.1-1.Only the Allowable Values are specified for each RPS trip Function in theLCO. Nominal trip setpoints are specified in the setpoint calculations.
Thenominal setpoints are selected to ensure that the setpoint measured byCHANNEL CALIBRATIONS does not exceed the Allowable Value. A tripsetpoint found less conservative than the nominal trip setpoint, but within itsAllowable Value, is considered OPERABLE with respect to the uncertainty allowances assumed for the applicable surveillance interval provided thatOCONEE UNITS 1, 2, & 3B 3.3.1 -1112/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE operation, testing and subsequent calibration are consistent with theSAFETY ANALYSES, assumptions of the setpoint calculations.
Each Allowable Value specified isLCO, and more conservative than instrument uncertainties appropriate to the tripAPPLICABILITY Function.
These uncertainties are defined in Reference 4.(continued)
For most RPS Functions, the Allowable Value in conjunction with thenominal trip setpoint ensure that the departure from nucleate boiling (DNB),center line fuel melt, or RCS pressure SLs are not challenged.
Cyclespecific values for use during operation are contained in the COLR.Certain RPS trips function to indirectly protect the SLs by detecting specificconditions that do not immediately challenge SLs but will eventually lead tochallenge if no action is taken. These trips function to minimize the unittransients caused by the specific conditions.
The Allowable Value for theseFunctions is selected at the minimum deviation from normal values that willindicate the condition, without risking spurious trips due to normalfluctuations in the measured parameter.
The safety analyses applicable to each RPS Function are discussed next.1. Nuclear Overpower
: a. Nuclear Overpower
-High SetpointThe Nuclear Overpower
-High Setpoint trip providesprotection for the design thermal overpower condition basedon the measured out of core neutron leakage flux.The Nuclear Overpower
-High Setpoint trip initiates a reactortrip when the neutron power reaches a predefined setpoint atthe design overpower limit. Because THERMAL POWER lagsthe neutron power, tripping when the neutron power reachesthe design overpower will limit THERMAL POWER to preventexceeding acceptable fuel damage limits.Thus, the Nuclear Overpower
-High Setpoint trip protectsagainst violation of the DNBR and fuel centerline melt SLs.However, the RCS Variable Low Pressure, and NuclearOverpower Flux/Flow Imbalance, provide more directprotection.
The role of the Nuclear Overpower
-High Setpointtrip is to limit reactor THERMAL POWER below the highestpower at which the other two trips are known to provideprotection.
The Nuclear Overpower
-High Setpoint trip also providestransient protection for rapid positive reactivity excursions OCONEE UNITS 1, 2, & 3B 3.3.1-1212/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE
: a. Nuclear Overpower
-High Setpoint (continued)
SAFETY ANALYSES, LCO, and during power operations.
These events include the rodAPPLICABILITY withdrawal accident and the rod ejection accident.
Byproviding a trip during these events, the Nuclear Overpower
-High Setpoint trip protects the unit from excessive power levelsand also serves to limit reactor power to prevent violation ofthe RCS pressure SL.Rod withdrawal accident analyses cover a large spectrum ofreactivity insertion rates (rod worths),
which exhibit slow andrapid rates of power increases.
At high reactivity insertion rates, the Nuclear Overpower
-High Setpoint trip provides theprimary protection.
At low reactivity insertion rates, the highpressure trip provides primary protection.
: b. Nuclear Overpower
-Low SetpointWhen initiating shutdown bypass, the Nuclear Overpower
-Low Setpoint trip must be reduced to _< 5% RTP. The lowpower setpoint, in conjunction with the lower Shutdown BypassRCS High Pressure
: setpoint, ensure that the unit is protected from excessive power conditions when other RPS trips arebypassed.
The setpoint Allowable Value was chosen to be as low aspractical and still lie within the range of the out of coreinstrumentation.
: 2. RCS High Outlet Temperature The RCS High Outlet Temperature trip, in conjunction with the RCSLow Pressure and RCS Variable Low Pressure trips, providesprotection for the DNBR SL. A trip is initiated whenever the reactorvessel outlet temperature approaches the conditions necessary forDNB. Portions of each RCS High Outlet Temperature trip channelare common with the RCS Variable Low Pressure trip. The RCSHigh Outlet Temperature trip provides steady state protection for theDNBR SL.The RCS High Outlet Temperature trip limits the maximum RCStemperature to below the highest value for which DNB protection bythe Variable Low Pressure trip is ensured.
The trip setpointAllowable Value is selected to ensure that a trip occurs before hot legtemperatures reach the point beyond which the RCS Low PressureOCONEE UNITS 1, 2, & 3B 3.3.1-1312/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE
: 2. RCS High Outlet Temperature (continued)
SAFETY ANALYSES, LCO, and and Variable Low Pressure trips are analyzed.
Above the highAPPLICABILITY temperature trip, the variable low pressure trip need not provideprotection, because the unit would have tripped already.
Thesetpoint Allowable Value does not reflect errors induced by harshenvironmental conditions that the equipment is expected toexperience because the trip is not required to mitigate accidents thatcreate harsh conditions in the RB.3. RCS High PressureThe RCS High Pressure trip works in conjunction with thepressurizer and main steam relief valves to prevent RCSoverpressurization, thereby protecting the RCS High Pressure SL.The RCS High Pressure trip has been credited in the transient analysis calculations for slow positive reactivity insertion transients (rod withdrawal transients and moderator dilution).
The rodwithdrawal transient covers a large spectrum of reactivity insertion rates and rod worths that exhibit slow and rapid rates of powerincreases.
At high reactivity insertion rates, the Nuclear Overpower
-High Setpoint trip provides the primary protection.
At low reactivity insertion rates, the RCS High Pressure trip provides the primaryprotection.
The setpoint Allowable Value is selected to ensure that the RCSHigh Pressure SL is not challenged during steady state operation orslow power increasing transients.
The setpoint Allowable Value doesnot reflect errors induced by harsh environmental conditions becausethe equipment is not required to mitigate accidents that create harshconditions in the RB.4. RCS Low PressureThe RCS Low Pressure trip, in conjunction with the RCS High OutletTemperature and Variable Low Pressure trips, provides protection for the DNBR SL. A trip is initiated whenever the system pressureapproaches the conditions necessary for DNB. The RCS LowPressure trip provides DNB low pressure limit for the RCS VariableLow Pressure trip.The RCS Low Pressure setpoint Allowable Value is selected toensure that a reactor trip occurs before RCS pressure is reducedbelow the lowest point at which the RCS Variable Low Pressure tripis analyzed.
The RCS Low Pressure trip provides protection forOCONEE UNITS 1, 2, & 3B 3.3.1-1412/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE
: 4. RCS Low Pressure (continued)
SAFETY ANALYSES, LCO, and primary system depressurization events and has been credited inAPPLICABILITY the accident analysis calculations for small break loss of coolantaccidents (LOCAs).
Harsh RB conditions created by small breakLOCAs cannot affect performance of the RCS pressure sensors andtransmitters within the time frame for a reactor trip. Therefore, degraded environmental conditions are not considered in theAllowable Value determination.
: 5. RCS Variable Low PressureThe RCS Variable Low Pressure trip, in conjunction with the RCSHigh Outlet Temperature and RCS Low Pressure trips, providesprotection for the DNBR SL. A trip is initiated whenever the systemparameters of pressure and temperature approach the conditions necessary for DNB. The RCS Variable Low Pressure trip provides afloating low pressure trip based on the RCS High OutletTemperature within the range specified by the RCS High OutletTemperature and RCS Low Pressure trips.The RCS Variable Low Pressure setpoint Allowable Value is selectedto ensure that a trip occurs when temperature and pressureapproach the conditions necessary for DNB while operating in atemperature pressure region constrained by the low pressure andhigh temperature trips. The RCS Variable Low Pressure trip isassumed for transient protection in the main steam line breakanalysis.
The setpoint allowable value does not include errorsinduced by the harsh environment, because the trip actuates prior tothe harsh environment.
: 6. Reactor Building High PressureThe Reactor Building High Pressure trip provides an early indication of a high energy line break (HELB) inside the RB. By detecting changes in the RB pressure, the RPS can provide a reactor tripbefore the other system parameters have varied significantly.
Thus,this trip acts to minimize accident consequences.
It also provides abackup for RPS trip instruments exposed to an RB HELBenvironment.
The Allowable Value for RB High Pressure trip is set at the lowestvalue consistent with avoiding spurious trips during normal operation.
The electronic components of the RB High Pressure trip are locatedin an area that is not exposed to high temperature steamenvironments during HELB transients inside containment.
TheOCONEE UNITS 1, 2, & 3B 3.3.1-1512/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE
: 6. Reactor Building High Pressure (continued)
SAFETY ANALYSES, LCO, and components are exposed to high radiation conditions.
Therefore, theAPPLICABILITY determination of the setpoint Allowable Value accounts for errorsinduced by the high radiation.
: 7. Reactor Coolant Pump to PowerThe Reactor Coolant Pump to Power trip provides protection forchanges in the reactor coolant flow due to the loss of multiple RCPs.Because the flow reduction lags loss of power indications due to theinertia of the RCPs, the trip initiates protective action earlier than atrip based on a measured flow signal.The Reactor Coolant Pump to Power trip has been credited in theaccident analysis calculations for the loss of more than two RCPs.The Allowable Value for the Reactor Coolant Pump to Power tripsetpoint is selected to prevent normal power operation unless atleast three RCPs are operating.
Each reactor coolant pump has anRCPPM, which monitors the electrical power and breaker status ofeach pump motor to determine if the pump is running.
EachRCPPM provides inputs to all four RPS channels.
The RCPPM willinitiate a reactor trip if fewer than three reactor coolant pumps areoperating and reactor power is greater than approximately 2%rated full power.8. Nuclear Overpower Flux/Flow Imbalance The Nuclear Overpower Flux/Flow Imbalance trip provides steadystate protection for the power imbalance SLs. A reactor trip isinitiated prior to the core power, AXIAL POWER IMBALANCE, andreactor coolant flow conditions exceeding the DNB or fuel centerline temperature limits.This trip supplements the protection provided by the Reactor CoolantPump to Power trip, through the power to flow ratio, for loss ofreactor coolant flow events. The power to flow ratio provides directprotection for the DNBR SL for the loss of one or more RCPs and forlocked RCP rotor accidents.
The power to flow ratio of the Nuclear Overpower Flux/Flow Imbalance trip also provides steady state protection to preventreactor power from exceeding the allowable power when the primarysystem flow rate is less than full four pump flow. Thus, the power toflow ratio prevents overpower conditions similar to the NuclearOCONEE UNITS 1, 2, & 3B 3.3.1-1612/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE
: 8. Nuclear Overpower Flux/Flow Imbalance (continued)
SAFETY ANALYSES, LCO, and Overpower trip. This protection ensures that during reduced flowAPPLICABILITY conditions the core power is maintained below that required to beginDNB.The Allowable Value is selected to ensure that a trip occurs when thecore power, axial power peaking, and reactor coolant flow conditions indicate an approach to DNB or fuel centerline temperature limits.By measuring reactor coolant flow and by tripping only whenconditions approach an SL, the unit can operate with the loss of onepump from a four pump initial condition at power levels at least aslow as approximately 80% RTP. The Allowable Value for theFunction, including the upper limits of the Function are given in theunit COLR because the cycle specific core peaking changes affectthe Allowable Value.9. Main Turbine Trip (Hydraulic Fluid Pressure)
The Main Turbine Trip Function trips the reactor when the mainturbine is lost at high power levels. The Main Turbine Trip Functionprovides an early reactor trip in anticipation of the loss of heat sinkassociated with a turbine trip. The Main Turbine Trip Function wasadded to the B&W designed units in accordance with NUREG-0737 (Ref. 5) following the Three Mile Island Unit 2 accident.
The triplowers the probability of an RCS power operated relief valve (PORV)actuation for turbine trip cases. This trip is activated at higher powerlevels, thereby limiting the range through which the Integrated Control System must provide an automatic runback on a turbine trip.Each of the four turbine hydraulic fluid pressure switches feeds oneprotective channel that continuously monitors the status of thecontacts.
For the Main Turbine Trip (Hydraulic Fluid Pressure),
the Allowable Value of 800 psig is selected to provide a trip whenever main turbinehydraulic fluid pressure drops below the normal operating range.This trip is bypassed at power levels < 30% RTP for unit startup.The turbine trip is not required to protect against events that cancreate a harsh environment in the turbine building.
Therefore, errorsinduced by harsh environments are not included in the determination of the setpoint Allowable Value.OCONEE UNITS 1, 2, & 3B 3.3.1-1712/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE SAFETY ANALYSES, LCO, andAPPLICABILITY (continued)
: 10. Loss of Main Feedwater PumD Turbines (Hydraulic Oil Pressure)
The Loss of Main Feedwater Pump Turbines (Hydraulic OilPressure) trip provides a reactor trip at high power levels when bothMFW pump turbines are lost. The trip provides an early reactor tripin anticipation of the loss of heat sink associated with the LOMF.This trip was added in accordance with NUREG-0737 (Ref. 5)following the Three Mile Island Unit 2 accident.
This trip provides areactor trip at high power levels for a LOMF to minimize challenges to the PORV.For the feedwater pump turbine hydraulic oil pressure, the Allowable Value of 75 psig is selected to provide a trip whenever feedwater pump turbine hydraulic oil pressure drops below the normaloperating range. This trip is bypassed at power levels < 2% RTP forunit startup.
The Loss of Main Feedwater Pump Turbines (Hydraulic Oil Pressure) trip is not required to protect against events that cancreate a harsh environment in the turbine building.
Therefore, errorscaused by harsh environments are not included in the determination of the setpoint Allowable Value.11. Shutdown Bypass RCS High PressureThe RPS Shutdown Bypass RCS High Pressure is provided to allowfor withdrawing the CONTROL RODS prior to reaching the normalRCS Low Pressure trip setpoint.
The shutdown bypass provides tripprotection during deboration and RCS heatup by allowing theoperator to at least partially withdraw the safety groups of CONTROLRODS. This makes their negative reactivity available to terminate inadvertent reactivity excursions.
Use of the shutdown bypass triprequires that the neutron power trip setpoint be reduced to 5% of fullpower or less. The Shutdown Bypass RCS High Pressure trip forcesa reactor trip to occur whenever the unit switches from poweroperation to shutdown bypass or vice versa. This ensures that theCONTROL RODS are all inserted before power operation can begin.The operator is required to remove the shutdown bypass, reset theNuclear Overpower
-High Power trip setpoint, and again withdrawthe safety group rods before proceeding with startup.Accidents analyzed in the UFSAR, Chapter 15 (Ref. 2), do notdescribe events that occur during shutdown bypass operation, because the consequences of these events are enveloped by theevents presented in the UFSAR.OCONEE UNITS 1, 2, & 3B 3.3.1-1812/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE SAFETY ANALYSES, LCO, andAPPLICABILITY
: 11. Shutdown Bypass RCS High Pressure (continued)
During shutdown bypass operation with the Shutdown Bypass RCSHigh Pressure trip active with a setpoint of < 1720 psig and theNuclear Overpower
-Low Setpoint set at or below 5% RTP, the tripslisted below can be bypassed.
Under these conditions, theShutdown Bypass RCS High Pressure trip and the NuclearOverpower
-Low Setpoint trip act to prevent unit conditions fromreaching a point where actuation of these Functions is necessary.
1 a. Nuclear Overpower
-High Setpoint;
: 3. RCS High Pressure;
: 4. RCS Low Pressure;
: 5. RCS Variable Low Pressure;
: 7. Reactor Coolant Pump to Power; and8. Nuclear Overpower Flux/Flow Imbalance.
The Shutdown Bypass RCS High Pressure Function's Allowable Value is selected to ensure a trip occurs before producing THERMALPOWER.General Discussion The RPS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 7). In MODES 1 and2, the following trips shall be OPERABLE because the reactor can becritical in these MODES. These trips are designed to take the reactorsubcritical to maintain the SLs during anticipated transients and to assistthe ESPS in providing acceptable consequences during accidents.
1 a. Nuclear Overpower
-High Setpoint;
: 2. RCS High Outlet Temperature;
: 3. RCS High Pressure;
: 4. RCS Low Pressure;
: 5. RCS Variable Low Pressure;
: 6. Reactor Building High Pressure; OCONEE UNITS 1, 2, & 3B 3.3.1-1912/10/14 1
RPS Instrumentation B 3.3.1BASESAPPLICABLE General Discussion (continued)
SAFETY ANALYSES, LCO, and 7. Reactor Coolant Pump to Power; andAPPLICABILITY
: 8. Nuclear Overpower Flux/Flow Imbalance.
Functions la, 3, 4, 5, 7, and 8 just listed may be bypassed in MODE 2when RCS pressure is below 1720 psig, provided the Shutdown BypassRCS High Pressure and the Nuclear Overpower
-Low setpoint trip areplaced in operation.
Under these conditions, the Shutdown Bypass RCSHigh Pressure trip and the Nuclear Overpower
-Low setpoint trip act toprevent unit conditions from reaching a point where actuation of theseFunctions is necessary.
The Main Turbine Trip (Hydraulic Fluid Pressure)
Function is required to beOPERABLE in MODE 1 at _ 30% RTP. The Loss of Main Feedwater Pump Turbines (Hydraulic Oil Pressure)
Function is required to beOPERABLE in MODE 1 and in MODE 2 at __ 2% RTP. For operation belowthese power levels, these trips are not necessary to minimize challenges tothe PORVs as required by NUREG-0737 (Ref. 5).Because the safety function of the RPS is to trip the CONTROL RODS, theRPS is not required to be OPERABLE in MODE 3, 4, or 5 if either thereactor trip breakers are open, or the CRD System is incapable of rodwithdrawal.
Similarly, the RPS is not required to be OPERABLE in MODE6 because the CONTROL RODS are normally decoupled from the CRDs.However, in MODE 2, 3, 4, or 5, the Shutdown Bypass RCS High Pressureand Nuclear Overpower
-Low setpoint trips are required to be OPERABLEif the CRD trip breakers are closed and the CRD System is capable of rodwithdrawal.
Under these conditions, the Shutdown Bypass RCS HighPressure and Nuclear Overpower
-Low setpoint trips are sufficient toprevent an approach to conditions that could challenge SLs.ACTIONS Conditions A and B are applicable to all RPS protective Functions.
If achannel's trip setpoint is found nonconservative with respect to the requiredAllowable Value in Table 3.3.1-1, or the transmitter, instrument loop, signalprocessing electronics or processor output trip device is found inoperable, the channel must be declared inoperable and Condition A enteredimmediately.
When an RPS channel is manually
: tripped, the functions that wereinoperable prior to tripping remain inoperable.
Other functions in the samechannel that were OPERABLE prior to tripping remain OPERABLE.
OCONEE UNITS 1, 2, & 3B 3.3.1-2012/10/14 1
RPS Instrumentation B 3.3.1BASESACTIONS A._1(continued)
For Required Action A.1, if one or more Functions in a required protective channel becomes inoperable, the affected protective channel must beplaced in trip.Placing the affected Function in trip places only the affected Function ineach required channel in a one-out-of-two logic configuration.
If thesame function in another channel exceeds the setpoint, all channels willtrip. In this configuration, the RPS can still perform its safety function inthe presence of a random failure of any single Channel.
The 4 hourCompletion Time is justified based on the continuous monitoring andsignal validation being performed and is sufficient time to place aFunction in trip. If the individual Function cannot be placed in trip, theOperator can trip the affected channel with the use of the Manual TripKeyswitch until such time that the Function can be placed in trip. Thisplaces all RPS Functions in a one-out-of-two logic configuration.
B.1Required Action B.1 directs entry into the appropriate Condition referenced in Table 3.3.1-1.
The applicable Condition referenced in the table isFunction dependent.
If the Required Action and the associated Completion Time of Condition A are not met or if more than two channels areinoperable, Condition B is entered to provide for transfer to the appropriate subsequent Condition.
C.1 and C.2If the Required Action and associated Completion Time of Condition A arenot met and Table 3.3.1-1 directs entry into Condition C, the unit must bebrought to a MODE in which the specified RPS trip Functions are notrequired to be OPERABLE.
The allowed Completion Time of 12 hours isreasonable, based on operating experience, to reach MODE 3 from fullpower conditions in an orderly manner and to open all CRD trip breakerswithout challenging unit systems.D..1If the Required Action and associated Completion Time of Condition A arenot met and Table 3.3.1-1 directs entry into Condition D, the unit must bebrought to a MODE in which the specified RPS trip Functions are notOCONEE UNITS 1, 2, & 3B 3.3.1-2112/10/14 1
RPS Instrumentation B 3.3.1BASESACTIONS D.1 (continued) required to be OPERABLE.
To achieve this status, all CRD tdp breakersmust be opened. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to open CRD tdp breakers withoutchallenging unit systems.E.1If the Required Action and associated Completion Time of Condition A arenot met and Table 3.3.1-1 directs entry into Condition E, the unit must bebrought to a MODE in which the specified RPS trip Function is not requiredto be OPERABLE.
To achieve this status, THERMAL POWER must bereduced < 30% RTP. The allowed Completion Time of 6 hours isreasonable, based on operating experience, to reach 30% RTP from fullpower conditions in an orderly manner without challenging unit systems.F.1If the Required Action and associated Completion Time of Condition A arenot met and Table 3.3.1-1 directs entry into Condition F, the unit must bebrought to a MODE in which the specified RPS trip Function is not requiredto be OPERABLE.
To achieve this status, THERMAL POWER must bereduced < 2% RTP. The allowed Completion Time of 12 hours isreasonable, based on operating experience, to reach 2% RTP from fullpower conditions in an orderly manner without challenging unit systems.SURVEILLANCE The SRs for each RPS Function are identified by the SRs column ofREQUIREMENTS Table 3.3.1-1 for that Function.
Most Functions are subject to CHANNELCHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION testing.The SRs are modified by a Note. The Note directs the reader to Table3.3.1-1 to determine the correct SRs to perform for each RPS Function.
SR 3.3.1.1Performance of the CHANNEL CHECK once every 12 hours ensures that agross failure of instrumentation has not occurred.
OCONEE UNITS 1, 2, & 3B 3.3.1-2212/10/14 1
RPS Instrumentation B 3.3.1BASESSURVEILLANCE SR 3.3.1.1 (continued)
REQUIREMENTS A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.
It is based onthe assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even moreserious.
CHANNEL CHECK will detect gross channel failure; therefore, itis key in verifying that the instrumentation continues to operate properlybetween each CHANNEL CALIBRATION.
Agreement criteria are determined based on a combination of the channelinstrument uncertainties, including isolation, indication, and readability.
If achannel is outside the criteria, it may be an indication that the transmitter orthe signal processing equipment has drifted outside its limit. If thechannels are within the criteria, it is an indication that the channels areOPERABLE.
If the channels are normally off scale during times whensurveillance is required, the CHANNEL CHECK will only verify that they areoff scale in the same direction.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.
The CHANNEL CHECKsupplements less formal but more frequent checks of channelOPERABILITY during normal operational use of the displays associated with the LCO's required channels.
For Functions that trip on a combination of several measurements, such asthe Nuclear Overpower Flux/Flow Imbalance
: Function, the CHANNELCHECK must be performed on each input.The CHANNEL CHECK requirement is met automatically.
The digitalRPS provides continuous online automatic monitoring of each of the inputsignals in each channel, performs signal online validation againstrequired acceptance
: criteria, and provides hardware functional validation.
If any protective channel input signal is identified to be in the failurestatus, this condition is alarmed on the Unit Statalarm and input to theplant OAC. Immediate notification of the failure status is provided to theOperations staff.OCONEE UNITS 1, 2, & 3B 3.3.1-2312/10/14 1
RPS Instrumentation B 3.3.1BASESSURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.2This SR is the performance of a heat balance calibration for the powerrange channels when reactor power is > 15% RTP. The heat balancecalibration consists of a comparison of the results of the calorimetric withthe power range channel output. The outputs of the power range channelsare normalized to the calorimetric.
If the calorimetric exceeds the NuclearInstrumentation System (NIS) channel output by > 2% RTP, the NIS is notdeclared inoperable but must be adjusted.
If the NIS channel cannot beproperly
: adjusted, the channel is declared inoperable.
A Note clarifies thatthis Surveillance is required to be performed only if reactor power is >_ 15%RTP and that 24 hours is allowed for performing the first Surveillance afterreaching 15% RTP. At lower power levels, calorimetric data are lessaccurate.
The power range channel's output shall be adjusted consistent with thecalorimetric results if the calorimetric exceeds the power range channel's output by _> 2% RTP. The value of 2% is adequate because this value isassumed in the safety analyses of UFSAR, Chapter 15 (Ref. 2). Thesechecks and, if necessary, the adjustment of the power range channelsensure that channel accuracy is maintained within the analyzed errormargins.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.3.1.3A comparison of power range nuclear instrumentation channels againstincore detectors shall be performed when reactor power is _> 15% RTP. ANote clarifies that 24 hours is allowed for performing the first Surveillance after reaching 15% RTP. If the absolute value of imbalance error is >_ 2%RTP, the power range channel is not inoperable, but an adjustment of themeasured imbalance to agree with the incore measurements is necessary.
The Imbalance error calculation is adjusted for conservatism by applying acorrelation slope (CS) value to the error calculation formula.
This ensuresthat the value of the APIo is > API,. The CS value is listed in the COLR andis cycle dependent.
If the power range channel cannot be properlyrecalibrated, the channel is declared inoperable.
The calculation of theAllowable Value envelope assumes a difference in out of core to incoremeasurements of 2.0%. Additional inaccuracies beyond those that aremeasured are also included in the setpoint envelope calculation.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.OCONEE UNITS 1, 2, & 3B 3.3.1-2412/10/14 1
RPS Instrumentation B 3.3.1BASESSURVEILLANCE SR 3.3.1.4REQUIREMENTS (continued)
This SR has been deleted.SR 3.3.1.5This SR manually retrieves the software setpoints and verifies they arecorrect.
The proper functioning of the processor portion of the channel iscontinuously checked by an automatic cyclic self monitoring.
Verification of field instrument setpoints is not required by this surveillance.
Thissurveillance does not apply to the Reactor Building Pressure Functionbecause it consists of pressure switches which provide a contact statusto the system and there is no software setpoint to verify.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.3.1.6This SR requires manual actuation of the output channel interposing relays to demonstrate OPERABILITY of the relays. The properfunctioning of the processor portion of the channel is continuously checked by an automatic cyclic self monitoring.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.3.1.7A Note to the Surveillance indicates that neutron detectors are excludedfrom CHANNEL CALIBRATION.
This Note is necessary because of thedifficulty in generating an appropriate detector input signal. Excluding thedetectors is acceptable because the principles of detector operation ensurevirtually instantaneous response.
A CHANNEL CALIBRATION is a complete check of the instrument
: channel, including the sensor. The test verifies that the channel respondsto the measured parameter within the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel adjusted to account forinstrument drift to ensure that the instrument channel remains operational between successive tests. CHANNEL CALIBRATION shall find thatOCONEE UNITS 1, 2, & 3B 3.3.1-2512/10/14 1
RPS Instrumentation B 3.3.1BASESSURVEILLANCE SR 3.3.1.7 (continued)
REQUIREMENTS measurement errors and processor output trip device setpoint errors arewithin the assumptions of the uncertainty analysis.
Whenever a sensingelement is replaced, the CHANNEL CALIBRATION of the resistance temperature detectors (RTD) sensors is accomplished by an inplace crosscalibration that compares the other sensing elements with the recentlyinstalled sensing element.Since the CHANNEL FUNCTIONAL TEST is a part of the CHANNELCALIBRATION a separate SR is not required.
The digital RPS softwareperforms a continuous online automated cross channel check, separately for each channel, and continuous online signal error detection andvalidation.
The protection system also performs continuous onlinehardware monitoring.
The CHANNEL CALIBRATION essentially validates the self monitoring function and checks for a small set of failuremodes that are undetectable by the self monitoring function.
The digital processors shall be rebooted as part of the calibration.
Thisverifies that the software has not changed.
Signals into the system (fromthe field instrument or at the protective system cabinet) are applied duringthe channel calibration to ensure that the instrumentation is within thespecified allowance requirements.
This, in combination with ensuring thesetpoints are entered into the software correctly per SR 3.3.1.5, verifiesthe setpoints are within the Allowable Values.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES
: 1. UFSAR, Chapter 7.2. UFSAR, Chapter 15.3. 10 CFR 50.49.4. EDM-1 02, "Instrument Setpoint/Uncertainty Calculations."
: 5. NUREG-0737, "Clarification of TMI Action Plan Requirements,"
November 1979.6. BAW-10167, May 1986.7. 10 CFR 50.36.OCONEE UNITS 1, 2, & 3B 3.3.1-2612/10/14 1
RPS -RTCB 3.3.3B 3.3 INSTRUMENTATION B 3.3.3 Reactor Protective System (RPS) -Reactor Trip Component (RTC)BASESBACKGROUND The RPS consists of four independent protection
: channels, each containing an RTC. Figure 7.1 of UFSAR, Chapter 7 (Ref. 1), shows a typical RPSprotection channel and the relationship of the RTC to the RPSinstrumentation, manual trip, and CONTROL ROD drive (CRD) tripdevices.The RTC is made up of two digital output modules and four Reactor TripRelays (RTR) all contained within the respective RPS channel's cabinet.The RTC receives a channel trip signal in its own channel and channeltrip signals from the digital output modules in the other three RPSchannels.
Whenever any two RPS channels transmit channel trip signals, the RTClogic in each channel actuates to remove 120 VAC power from itsassociated CRD trip devices.The RPS trip scheme consists of processor output trip devices.During normal unit operations, the digital output modules maintain theRTRs energized.
: However, if an RPS channel initiates a trip signal, thedigital output modules in that RPS channel will de-energize the reactortrip relay in that RPS channel and the associated RTR in each of theother three RPS channels.
When an RPS channel provides a trip signal, the digital output modules inthat RPS channel de-energize RTRs such that the following occurs:a. Each of the four (4) RTRs driven by that RPS channel's digitaloutput modules "informs" its associated RPS channel that areactor trip signal has occurred in the tripped RPS channel;b. The contacts in the trip device circuitry, powered by the trippedchannel, open, but the trip device remains energized through theclosed contacts from the RTRs of the other RTCs. (This condition exists in each RPS -RTC. Each RPS -RTC controls power to atrip device.)OCONEE UNITS 1, 2, & 3B 3.3.3-112/10/14 1
RPS -RTCB 3.3.3BASESBACKGROUND (continued)
When the second RPS channel senses a reactor trip condition, the RTRsdriven by the digital output modules for the second channel de-energize and open contacts that supply power to the trip devices.
With contactsopened by two separate RPS channels, power to the trip devices isinterrupted and the CONTROL RODS fall into the core.A minimum of two out of four RTCs must sense a trip condition to cause areactor trip.Because of the interchannel communication and 2.MIN/2.MAX (foranalog inputs) and two-out-of-four (for binary inputs),
an RPS channel willnot provide a trip signal to its RTC until trip conditions are satisfied in atleast two RPS channels for the same trip function.
The contacts of the four reactor trip relays in each RPS Channel cabinetare wired in a two-out-of-four logic scheme. The relays de-energize tode-energize the Control Rod Drive Breaker undervoltage circuit wired tothat channel and cause the shunt trip coil monitoring the circuit to beenergized.
Either de-energizing the undervoltage circuit or energizing theshunt trip circuit trips the CRD breaker.IAPPLICABLE Transient and accident analyses rely on a reactor trip for protection ofSAFETY ANALYSES reactor core integrity, reactor coolant pressure boundary integrity, andreactor building OPERABILITY.
A reactor trip must occur when needed toprevent accident conditions from exceeding those calculated in theaccident analyses.
More detailed descriptions of the applicable accidentanalyses are found in the bases for each of the RPS trip Functions in LCO3.3.1, "Reactor Protective System (RPS) Instrumentation."
The RTCs satisfy Criterion 3 of 10 CFR 50.36 (Ref. 2).LCO LCO 3.3.3 requires all four RTCs to be OPERABLE.
Failure of any RTCrenders a portion of the RPS inoperable.
An OPERABLE RTC must be able to receive and interpret trip signals fromOPERABLE RPS channels and to open its associated trip device.The requirement of four RTCs to be OPERABLE ensures that a minimum oftwo RTCs will remain OPERABLE if a single failure has occurred in one RTCand if a second RTC is out of service.
This two-out-of-four trip logic alsoensures that a single RTC failure will not cause an unwanted reactor trip.Violation of this LCO could result in a trip signal not causing a reactor tripwhen needed.IOCONEE UNITS 1, 2, & 3B 3.3.3-212/10/14 1
RPS -RTCB 3.3.3BASES (continued)
APPLICABILITY The RTCs are required to be OPERABLE in MODES 1 and 2. They arealso required to be OPERABLE in MODES 3, 4, and 5 if any CRD tripbreakers are in the closed position and the CRD System is capable of rodwithdrawal.
The RTCs are designed to ensure a reactor trip would occur, ifneeded. This condition can exist in all of these MODES; therefore, theRTCs must be OPERABLE.
ACTIONS A.1 and A.2When an RTC is inoperable, the associated CRD trip breaker must then beplaced in a condition that is equivalent to a tripped condition for the RTC.Required Action A.1 or Required Action A.2 requires this either by trippingthe CRD trip breaker or by removing power to the CRD trip device. Trippingone RTC or removing power opens one of the CRD trip devices, which willresult in the loss of one of the parallel power supplies.
Power to holdCONTROL RODS in position is still provided via the parallel CRD powersupply. Therefore, a reactor trip will not occur until a second protection channel trips.B.1, B.2.1. and B.2.2Condition B applies if two or more RTCs are inoperable or if the RequiredAction and associated Completion Time of Condition A are not met inMODE 1, 2, or 3. In this case, the unit must be placed in a MODE in whichthe LCO does not apply. This is done by placing the unit in at least MODE3 with all CRD trip breakers open or with power from all CRD trip breakersremoved within 12 hours. The allowed Completion Time of 12 hours isreasonable, based on operating experience, to reach MODE 3 from fullpower conditions in an orderly manner and without challenging unitsystems.C.1 and C.2Condition C applies if two or more RTCs are inoperable or if the RequiredAction and associated Completion Time of Condition A are not met inMODE 4 or 5. In this case, the unit must be placed in a MODE in which theLCO does not apply. This is done by opening all CRD trip breakers orremoving power from all CRD trip breakers.
The allowed Completion Timeof 6 hours is reasonable, based on operating experience, to open all CRDtrip breakers or remove power from all CRD trip breakers withoutchallenging unit systems.OCONEE UNITS 1, 2, & 3B 3.3.3-3i-2/1-0/14 I
RPS -RTCB 3.3.3BASES (continued)
SURVEILLANCE REQUIREMENTS SR 3.3.3.1The SRs include performance of a CHANNEL FUNCTIONAL TEST. Thistest shall verify the OPERABILITY of the RTC and its ability to receive andproperly respond to channel trip and reactor trip signals.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.This testing is normally performed on a rotational basis. Testing in thismanner reduces the likelihood of the same systematic test errors beingintroduced into each redundant RTC.REFERENCES
: 1. UFSAR, Chapter 7.2. 10 CFR 50.36.OCONEE UNITS 1, 2, & 3B 3.3.3-412/10/14 1
CRD Trip DevicesB 3.3.4B 3.3 INSTRUMENTATION B 3.3.4 Control Rod Drive (CRD) Trip DevicesBASESBACKGROUND The Reactor Protective System (RPS) contains multiple CRD trip devicesin the form of four AC trip breakers.
The system has two separate paths(or channels),
with each path having two AC breakers in series. In eithercase, each path provides independent power to the CRDs. Also, in eithercase, either path can provide sufficient power to operate the entire CRDSystem.Figure 7.1, UFSAR, Chapter 7 (Ref. 1), illustrates the configuration ofReactor Protection System (RPS) Reactor Trip Components (RTC's) andthe trip breakers.
To trip the reactor, power to the CRDs must beremoved.
Loss of power causes the CRD mechanisms to release theCONTROL RODS, which then fall by gravity into the core.Power to CRIs is supplied from two separate sources through the AC tripcircuit breakers.
These breakers are designated A, B, C, and D. Theirundervoltage (trip) coils are powered by RPS channels A, B, C, and D,respectively and their shunt (trip) coils are actuated by RPS channels A,B, C, and D, respectively.
From the circuit breakers, the CRD powertravels through voltage regulators and stepdown transformers.
Thesedevices in turn supply redundant buses that feed the Single Rod PowerSupplies (SRPS).Two AC breakers (A and C) are in series to feed one redundant train of theSRPS, whereas the other two series AC breakers (B and D) feed the otherredundant train of the SRPS. The minimum required logic required tocause a reactor trip is the opening of a circuit breaker in each parallel pathto the SRPS. This is known as a one-out-of-two taken twice logic. Thefollowing examples illustrate the operation of the reactor trip circuitbreakers.
: a. If the A or C circuit breaker opens, input power to one train of theSRPS's is lost.b. If in addition, the B or D circuit breaker opens, input power to the othertrain of the SRPS's is lost, which will result in the dropping of all rods(except APSR's) into the core.OCONEE UNITS 1, 2, & 3B 3.3.4-112/10/14 1
CRD Trip DevicesB 3.3.4BASESBACKGROUND (continued)
The reactor trip relays located in RPS Channel A cabinet provide thetwo-out-of-four relay logic to trip CRD breaker A, relays in RPS B cabinettrip CRD breaker B, relays in RPS C cabinet trip CRD breaker C, andrelays in RPS D cabinet trip CRD breaker D. If two or more channels ofRPS indicate a valid software trip logic condition (two-out-of-four),
thebinary outputs will de-energize the trip relays associated with thosechannels in all RPS cabinets, tripping all four CRD breakers resulting in areactor trip.IAPPLICABLE SAFETY ANALYSESAccident analyses rely on a reactor trip for protection of reactor coreintegrity, reactor coolant pressure boundary integrity, and reactor buildingOPERABILITY.
A reactor trip must occur when needed to preventaccident consequences from exceeding those calculated in the accidentanalyses.
The CONTROL ROD position limits ensure that adequate rodworth is available upon reactor trip to shut down the reactor to therequired SDM. Further, OPERABILITY of the CRD trip devices ensuresthat all CONTROL RODS will trip when required.
More detaileddescriptions of the applicable accident analyses are found in the Bases foreach of the individual RPS trip Functions in LCO 3.3.1, "Reactor Protective System (RPS) Instrumentation."
The CRD trip devices satisfy Criterion 3 of CFR 50.36 (Ref. 2).LCOThe LCO requires all of the specified CRD trip devices to be OPERABLE.
Failure of any required CRD trip device renders a portion of the RPSinoperable and reduces the reliability of the affected Functions.
Withoutreliable CRD reactor trip circuit breakers and associated support circuitry, areactor trip may not reliably occur when initiated either automatically ormanually.
All required CRD trip devices shall be OPERABLE to ensure that thereactor remains capable of being tripped any time it is critical.
OPERABILITY is defined as the CRD trip device being able to receive areactor trip signal and to respond to this trip signal by interrupting power tothe CRDs. Both of the CRD trip breaker's diverse trip devices and thebreaker itself must be functioning properly for the breaker to beOPERABLE.
Requiring all breakers to be OPERABLE ensures that at least one device ineach of the two power paths to the CRDs will remain OPERABLE evenwith a single failure.OCONEE UNITS 1, 2, & 3B83.3.4-2 12/10/14 1
CRD Trip DevicesB 3.3.4BASES (continued)
APPLICABILITY The CRD trip devices shall be OPERABLE in MODES 1 and 2, and inMODES 3, 4, and 5 when any CRD trip breaker is in the closed positionand the CRD System is capable of rod withdrawal.
The CRD trip devices are designed to ensure that a reactor trip wouldoccur if needed. Since this condition can exist in all of these MODES, theCRD trip devices shall be OPERABLE.
ACTIONS A Note has been added to the ACTIONS indicating separate Condition entry is allowed for each CRD trip device.A.1 and A.2Condition A represents reduced redundancy in the CRD trip Function.
Condition A applies when one diverse trip Function (undervoltage or shunttrip device) is inoperable in one or more CRD trip breaker(s).
If one of the diverse trip Functions on a CRD trip breaker becomesinoperable, actions must be taken to preclude the inoperable CRD tripdevice from preventing a reactor trip when needed. This is done bymanually tripping the inoperable CRD trip breaker or by removing powerfrom the inoperable CRD trip breaker.
Either of these actions places theaffected CRDs in a one-out-of-two trip configuration, which precludes asingle failure from preventing a reactor trip. The 48 hour Completion Timehas been shown to be acceptable through operating experience.
B.1 and B.2Condition B represents a loss of redundancy for the CRD trip Function.
Condition B applies when both diverse trip Functions are inoperable in oneor more trip breaker(s).
Required Action B.1 and Required Action B.2 are the same as RequiredAction A.1 and Required Action A.2, but the Completion Time is shortened.
The 1 hour Completion Time allowed to trip or remove power from the CRDtrip breaker allows the operator to take all the appropriate actions for theinoperable breaker and still ensures that the risk involved is acceptable.
OCONEE UNITS 1, 2, & 3B 3.3.4-312/10/14 1
CRD Trip DevicesB 3.3.4BASESACTIONS(continued)
C.1, C.2.1, and C.2.2With the Required Action and associated Completion Time of Condition Aor B not met in MODE 1, 2, or 3, the unit must be brought to a MODE inwhich the LCO does not apply. To achieve this status, the unit must bebrought to MODE 3, with all CRD trip breakers open or with power from allCRD trip breakers removed within 12 hours. The allowed Completion Timeof 12 hours is reasonable, based on operating experience, to reach MODE3 from full power conditions in an orderly manner and without challenging unit systems.D.1 and D.2With the Required Action and associated Completion Time of Condition Aor B not met in MODE 4 or 5, the unit must be brought to a MODE in whichthe LCO does not apply. To achieve this status, all CRD trip breakers mustbe opened or power from all CRD trip breakers removed within 6 hours.The allowed Completion Time of 6 hours is reasonable, based on operating experience, to open all CRD trip breakers or remove power from all CRDtrip breakers without challenging unit systems.SURVEILLANCE SR 3.3.4.1REQUIREMENTS SR 3.3.4.1 is to perform a CHANNEL FUNCTIONAL TEST. This testverifies the OPERABILITY of the trip devices by actuation of the enddevices.
Also, this test independently verifies the undervoltage and shunttrip mechanisms of the trip breakers.
The Surveillance Frequency is basedon operating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.REFERENCES
: 1. UFSAR, Chapter 7.2. 10 CFR 50.36.OCONEE UNITS 1, 2, & 3B 3.3.4-412/10/14 1
ESPS Input Instrumentation B 3.3.5B 3.3 INSTRUMENTATION B 3.3.5 Engineered Safeguards Protective System (ESPS) Input Instrumentation BASESBACKGROUND The ESPS initiates necessary safety systems, based on the values ofselected unit Parameters, to protect against violating core design limits andto mitigate accidents.
IESPS actuates the following systems:* High Pressure Injection (HPI);* Low Pressure Injection (LPI);* Reactor Building (RB) cooling;* RB Spray;0 RB Isolation; and* Keowee Hydro Unit Emergency Start.The ESPS operates in a distributed manner to initiate the appropriate systems.
The ESPS does this by determining the need for actuation ineach of three input channels monitoring each actuation Parameter.
Oncethe need for actuation is determined, the condition is transmitted toautomatic actuation output logic channels, which perform the two-out-of-three logic to determine the actuation of each end device. Each end devicehas its own automatic actuation logic, although all automatic actuation output logic channels take their signals from the same processor output tripdevice in each channel for each Parameter.
Four Parameters are used for actuation:
* Low Reactor Coolant System (RCS) Pressure; 0 Low Low RCS Pressure;
* High RB Pressure; and* High High RB Pressure.
IOCONEE UNITS 1, 2, & 3B 3.3.5-112/10/14 1
ESPS Input Instrumentation B 3.3.5BASESBACKGROUND LCO 3.3.5 covers only the input instrumentation channels that measure(continued) these Parameters.
These channels include all intervening equipment necessary to produce actuation before the measured process Parameter exceeds the limits assumed by the accident analysis.
This includessensors, processor output trip devices, operational bypass circuitry, andvoter input. LCO 3.3.6, "Engineered Safeguards Protective System(ESPS) Manual Initiation,"
and LCO 3.3.7, "Engineered Safeguards Protective System (ESPS) Automatic Actuation Output Logic Channels,"
provide requirements on the manual initiation and automatic actuation output logic Functions.
There are three input channels.
The ESPS Protective Channels A, B andC are made up of two independent subsystems
-one subsystem isinstalled in the ESPS cabinets and is designated A2, B2, and C2. Theother independent and redundant subsystem is installed in the RPScabinets and is designated Al, B1, and Cl. This subsystem uses theRPS protective channels (A, B, and C) computers.
The ESPS inputsignals are not redundant for the two subsystems.
The same inputsignals are fed to ESPS subsystems 1 and 2. The ESPS subsystems arefully redundant with the exception of the shared inputs. Each of thesetwo independent ESPS subsystems is fully capable of performing allrequired protective actions.The three ESPS channel computers in each subsystem areinterconnected via fiber optic data links, in a way that enables theexchange of data and signal online validation, before the calculation oftrip functions.
If the setpoint for a single input channel (for example, theRB High pressure input to Channel A) is exceeded, a channel tripstatalarm is actuated but a channel trip signal is not sent to the automatic actuation output logic channel.
Since the two ES subsystems shareinputs, this condition will be sensed by both Channel Al and A2. Also,due to the inter-channel communication, all 3 ES channels in eachsubsystem recognize that this input channel setpoint has been exceededfor one channel.
: However, due to the 2.MIN/2.MAX logic within thesystem, the same input channel setpoint for one of the other threechannels must be exceeded before channel trip signals are sent to theautomatic actuation output logic channels.
Again, due to the inter-channel communication, all 3 ES channels will then generate trip signalssince the 2.MIN/2.MAX condition has been satisfied.
The ESPS outputactuation signals are sent from ESPS protective channels A, B and C tothe ESPS actuation computers (Voters) via fiber optic data links. Figure7.5 UFSAR, Chapter 7 (Ref. 1), illustrates how input instrumentation channel trips combine to cause automatic actuation output logic channeltrips.OCONEE UNITS 1, 2, & 3B 3.3.5-212/10/14 1
ESPS Input Instrumentation B 3.3.5.BASESBACKGROUND (continued)
The following matrix identifies the input instrumentation (measurement) channels and the Automatic Actuation Output Logic Channels actuated byeach.Output Actuated RCS RCS RB RBLogic Channels Systems/
PRESS PRESS PRESS PRESSFunctions LOW LOW HIGH HIGHLOW HIGH1 and 2 HPI and RB Non-Essential X XIsolation, KeoweeEmergency Start, Load Shedand Standby Breaker Input,and Keowee Standby BusFeeder Breaker Input3 and 4 LPI X X5 and 6 RB Cooling and RB XEssential isolation 7 and 8 RB Spray XThe ES equipment is generally divided between the two redundant actuation output logic channels.
The division of the equipment between thetwo actuation output logic channels is based on the equipment redundancy and function and is accomplished in such a manner that the failure of oneof the actuation output logic channels and the related safeguards equipment will not inhibit the overall ES Functions.
Redundant ES pumpsare controlled from separate and independent actuation output logicchannels with some exceptions (e.g., HPI B pump which is actuated byboth).The actuation of ES equipment is also available by manual actuation switches located on the control room console.The ESPS, in conjunction with the actuated equipment, provides protective functions necessary to mitigate accidents, specifically the loss of coolantaccident (LOCA) and main steam line break (MSLB) events. The ESPSrelies on the OPERABILITY of the automatic actuation output logic for eachcomponent to perform the actuation of the selected systems of LCO 3.3.7.OCONEE UNITS 1, 2, & 3B 3.3.5-312/10/14 1
ESPS Input Instrumentation B 3.3.5BASESBACKGROUND Engineered Safeguards Protective System Bypasses(continued)
No provisions are made for maintenance bypass of ESPS instrumentation channels.
Operational bypass of certain input parameters is necessary toallow accident recovery actions to continue and, for some inputparameters, to allow unit shutdown without spurious ESPS actuation.
The ESPS RCS pressure instrumentation channel design allows ManualBypass when reactor pressure is below the point at which the low andlow low pressure trips are required to be OPERABLE.
Once permissive conditions are sensed, the RCS pressure trips may be manually bypassed.
Bypasses are automatically removed when bypass permissive conditions are exceeded.
This bypass provides an operational provision only outsidethe Applicability for this parameter, and provides no safety function.
There are two redundant subsystems.
The same input signal is fed toeach subsystem.
In subsystem 1, channels Al, B1, and Cl provide theinput to Voter 1 Odd and Voter 1 Even. In subsystem 2, channels A2,B2, and C2 provide input to Voter 2 Odd and Voter 2 Even. Eithersubsystem provides the full complement of Voters. This allows for aManual (maintenance)
Bypass of one complete subsystem, or portion ofa subsystem, without entering into an LCO Condition.
Parameter Change Enable ModeThe ESPS Instrument Input Channel A2, B2, and C2 processors caneach be placed in different operating modes through the use of the"Parameter Change Enable" keyswitches and commands from theService Unit. Each protective channel A2, B2,and C2 has a keyswitch located in that channel's cabinet pair.Placing ESPS Channels Al, B1 or Cl in Parameter Change EnableMode through the use of the "Parameter Change Enable" keyswitch located in the corresponding RPS cabinet will also place thecorresponding RPS Channels A, B, or C in Parameter Change EnableMode.When a keyswitch is placed from the normal Operating Mode position tothe Parameter Change Enable Mode position:
* The processors continue with normal operation.
* A permissive is provided that allows the Service Unit to be used tochange the operating mode of the processors associated with thatkeyswitch.
OCONEE UNITS 1, 2, & 3B 3.3.5-412/10/14 1
ESPS Input Instrumentation B 3.3.5BASESBACKGROUND Parameter Change Enable Mode (continued)
With the keyswitch in the Parameter Change Enable Position thefollowing modes of operation are allowed for processors:
* Normal Operation
-with permissive for operating mode change.* Parameterization
-allows changes to specific parameters (exampleplacing a parameter into a tripped condition or performing Go/NoGotesting).
* Function Test -for disabling the application function and forcingoutput signal for testing purposes (normally not used).* Diagnostics
-for downloading new application software.
The Function Test and Diagnostics modes result in the processor ceasingits cyclic processing of the application functions.
Entry into these modesfirst requires entry into Parameterization mode and setting a separateparameter.
When a keyswitch is placed in the Parameter Change Enable ModePosition for any activity, the affected processor shall first be declared outof service.
In addition to declaring the processor out of service, whenloading or revising software in a processor, the affected ESPS input shallbe tripped OR the associated ESPS voters shall be placed in Bypass. Ifthis activity is being performed on an ES Input Channel in subsystem 1,the associated RPS channel shall also be placed in manual bypass. Onlyone ESPS channel at a time is allowed to be placed into Parameter Change Enable Mode Position for software loading/revision.
Each Parameter Change Enable keyswitch status information is sent tothe Statalarm panel and to the OAC via the TXS Gateway.ESPS Parameter Change Enable keyswitches are administratively controlled (there are no hardware or software interlocks betweenchannels).
Reactor Coolant System PressureThe RCS pressure is monitored by three independent pressure transmitters located in the RB. These transmitters are separate from the transmitters that provide inputs to the Reactor Protective System (RPS). The output ofeach transmitter terminates in an input isolation module in the ESPS,which provides individually isolated output pressure signals.
Each of thepressure signals generated by these transmitters is monitored by twoindependent digital processing
: systems, with three ESPS input logicchannels and three RPS/ESPS input logic channels to provide two tripOCONEE UNITS 1, 2, & 3B 3.3.5-512/10/14 1
ESPS Input Instrumentation B 3.3.5BASESBACKGROUND Reactor Coolant System Pressure (continued)
: signals, at,>_ 1590 psig and _ 500 psig, and two bypass permissive signals,at _< 1750 psig and _< 900 psig.The outputs of the three logic processor channels in each of the twoprocessing subsystems (ESPS and RPS/ESPS) generate an output tripsignal to its associated independent actuation train Voters (Odd andEven) when the second minimum pressure signal of any of the threeinput channels falls below the Low RCS pressure setpoint.
This willinitiate an actuation of the Voter Output Channels 1 and 2 (HPIActuation).
The outputs of the input logic processors in each processing system also generate an output trip signal to its associated independent actuation train Voters (Odd and Even) when the second minimumpressure signal of the three input channels falls below the Low Low RCSpressure setpoint.
This will initiate an actuation of the Voter OutputChannels 3 and 4 (LPI Actuation).
Reactor Building PressureThere are three independent RB pressure transmitters.
The outputs of thethree logic processor channels in each of the two processing subsystems (ESPS and RPS/ESPS) generate an output trip signal to its associated independent actuation train Voters (Odd and Even) when the secondmaximum pressure signal of any of the three input channels increases above the High RB pressure setpoint.
This will initiate an actuation of VoterOutput Channels 5 and 6 (RB Cooling Actuation and RB Essential Isolation).
The outputs of the three high RB pressure processor output tripdevices also trip Voter Output Channels 1, 2, 3 and 4 to initiate HPI andLPI.The ESPS channels of the RB Spray System are formed by two separatetwo-out-of-three logic networks with the active elements originating in sixRB pressure sensing pressure switches.
One two-out-of-three networkactuates Channel 7 and the other two-out-of-three network actuatesChannel 8. Either of the two networks is capable of initiating the requiredprotective action.Trip Setpoints and Allowable ValuesTrip setpoints are the nominal value at which the processor output tripdevices are set. Any processor output trip device is considered to beproperly adjusted when the "as left" value is within the band forCHANNEL CALIBRATION accuracy.
OCONEE UNITS 1, 2, & 3B 3.3.5-612/10/14 1
ESPS Input Instrumentation B 3.3.5BASESBACKGROUND Trip Setpoints and Allowable Values (continued)
The trip setpoints used in the processor output trip devices are selectedsuch that adequate protection is provided when all sensor and processing time delays are taken into account.
To allow for calibration tolerances, instrumentation uncertainties, instrument drift, and severe environment induced errors for those ESPS channels that must function in harshenvironments as defined by 10 CFR 50.49 (Ref. 2), the Allowable Valuesspecified in Table 3.3.5-1 in the accompanying LCO are conservatively adjusted with respect to the analytical limits. A detailed description of themethodology used to calculate the trip setpoints and associated uncertainties is provided in Reference
: 3. The actual trip setpoint enteredinto the processor output trip device is more conservative than thatspecified by the Allowable Value to account for changes in randommeasurement errors detectable by a CHANNEL CALIBRATION.
Achannel is inoperable if its actual trip setpoint is not within its requiredAllowable Value.Setpoints, in accordance with the Allowable Values, ensure that theconsequences of accidents will be acceptable, providing the unit isoperated from within the LCOs at the onset of the accident and theequipment functions as designed.
With the exception of Reactor Building Pressure
-High High function, each channel is tested online by manually retrieving the software setpoint Ito ensure it has been entered correctly.
Signals into the system (from thefield instrument or at the protective system cabinet) are applied during thechannel calibration to ensure that the instrumentation is within thespecified allowance requirements.
The Reactor Building Pressure
-HighHigh actuation channel does not have software setpoints; it is actuated bya pressure switch that provides contact status only.APPLICABLE The following ESPS Functions have been assumed within the accidentSAFETY ANALYSES analyses.
High Pressure Iniection The ESPS actuation of HPI has been assumed for core cooling in theLOCA analysis and is credited with boron addition in the MSLB analysis.
Low Pressure Iniection The ESPS actuation of LPI has been assumed for large break LOCAs.OCONEE UNITS 1, 2, & 3B 3.3.5-712/10/14 1
ESPS Input Instrumentation B 3.3.5BASESAPPLICABLE Reactor Buildinq Spray, Reactor Buildinq
: Cooling, andSAFETY ANALYSES Reactor Building Isolation (continued)
The ESPS actuation of the RB coolers and RB Spray have been creditedin RB analysis for LOCAs, both for RB performance and equipment environmental qualification pressure and temperature envelope definition.
Accident dose calculations have credited RB Isolation and RB Spray.Keowee Hydro Unit Emergency StartThe ESPS initiated Keowee Hydro Unit Emergency Start has beenincluded in the design to ensure that emergency power is available throughout the limiting LOCA scenarios.
The small break LOCA analyses assume a conservative 48 second delaytime for the actuation of HPI and LPI in UFSAR, Chapter 15 (Ref. 4). Thelarge break LOCA analyses assume LPI flow starts in 38 seconds whilefull LPI flow does not occur until 36 seconds later, or 74 seconds total(Ref. 4). This delay time includes allowances for Keowee Hydro Unitstarting, Emergency Core Cooling Systems (ECCS) pump starts, andvalve openings.
Similarly, the RB Cooling, RB Isolation, and RB Sprayhave been analyzed with delays appropriate for the entire systemanalyzed.
Accident analyses rely on automatic ESPS actuation for protection of thecore temperature and containment pressure limits and for limiting off sitedose levels following an accident.
These include LOCA, and MSLBevents that result in RCS inventory reduction or severe loss of RCScooling.The ESPS channels satisfy Criterion 3 of 10 CFR 50.36 (Ref. 5).LCO The LCO requires three input channels of ESPS instrumentation for eachParameter in Table 3.3.5-1 to be OPERABLE in each ESPS automatic actuation output logic channel.
Failure of any instrument renders theaffected input channel(s) inoperable and reduces the reliability of theaffected Functions.
There are two redundant ESPS subsystems eachhaving three input channels.
Only one subsystem is required to beOPERABLE.
OCONEE UNITS 1, 2, & 3B 3.3.5-812/10/14 1
ESPS Input Instrumentation B 3.3.5BASESLCO(continued)
Only the Allowable Value is specified for each ESPS Function in theLCO. Nominal trip setpoints are specified in the setpoint calculations.
Thenominal trip setpoints are selected to ensure the setpoints measured byCHANNEL FUNCTIONAL TESTS or CHANNEL CALIBRATIONS do notexceed the Allowable Value if the processor output trip device is performing as required.
Operation with a trip setpoint less conservative than thenominal trip setpoint, but within its Allowable Value, is acceptable providedthat operation and testing are consistent with the assumptions of thesetpoint calculations.
Each Allowable Value specified is more conservative than the analytical limit assumed in the safety analysis to account forinstrument uncertainties appropriate to the trip Parameter.
Theseuncertainties are defined in Reference 3.The values for operating bypass removal functions are stated in theApplicable MODES or Other Specified Condition column of Table 3.3.5-1.Three ESPS input instrumentation channels shall be OPERABLE toensure that a single failure in one input channel will not result in loss ofthe ability to automatically actuate the required safety systems.The bases for the LCO on ESPS Parameters include the following.
Three input channels of RCS Pressure-Low, RCS Pressure-Low Low,RB Pressure-High and RB Pressure-High High are requiredOPERABLE.
Each channel includes a sensor, input isolation modules,interchannel communication modules and processor output trip devices.Failures that affect the ability to bypass an input channel do not renderthe input channel inoperable since the input channel is still capable ofperforming its safety function, i.e., this is not a safety related bypassfunction.
APPLICABILITY Three input channels of ESPS instrumentation for each of the following Parameters shall be OPERABLE.
: 1. Reactor Coolant System Pressure
-LowThe RCS Pressure
-Low actuation Parameter shall beOPERABLE during operation at or above 1750 psig. Thisrequirement ensures the capability to automatically actuate safetysystems and components during conditions indicative of a LOCAor secondary unit overcooling.
Below 1750 psig, the low RCSPressure actuation Parameter can be bypassed to avoid actuation during normal unit cooldowns when safety systems actuations arenot required.
OCONEE UNITS 1, 2, & 3B 3.3.5-912/10/14 1
ESPS Input Instrumentation B 3.3.5BASESAPPLICABILITY
: 1. Reactor Coolant System Pressure
-Low (continued)
The allowance for the bypass is consistent with the transition ofthe unit to a lower energy state, providing greater margins tosafety limits. The unit response to any event, given that thereactor is already tripped, will be less severe and allows sufficient time for operator action to provide manual safety systemactuations.
This is even more appropriate during unit heatupswhen the primary system and core energy content is low, prior topower operation.
In MODES 5 and 6, there is adequate time for the operator toevaluate unit conditions and respond by manually startingindividual
: systems, pumps, and other equipment to mitigate theconsequences of an abnormal condition or accident.
RCSpressure and temperature are very low, and many EScomponents are administratively controlled or otherwise prevented from actuating to prevent inadvertent overpressurization of unit systems.2. Reactor Coolant System Pressure
-Low LowThe RCS Pressure
-Low Low actuation Parameter shall beOPERABLE during operation above 900 psig. This requirement ensures the capability to automatically actuate safety systems andcomponents during conditions indicative of a LOCA or secondary unit overcooling.
Below 900 psig, the low low RCS Pressureactuation Parameter can be bypassed to avoid actuation duringnormal unit cooldowns when safety system actuations are notrequired.
The allowance for the bypass is consistent with the transition ofthe unit to a lower energy state, providing greater margins tosafety limits. The unit response to any event, given that thereactor is already tripped, will be less severe and allows sufficient time for operator action to provide manual safety systemactuations.
This is even more appropriate during unit heatupswhen the primary system and core energy content is low, prior topower operation.
OCONEE UNITS 1, 2, & 3B 3.3.5-1012/10/14 1
ESPS Input Instrumentation B 3.3.5BASESAPPLICABILITY
: 2. Reactor Coolant System Pressure
-Low Low (continued)
In MODES 5 and 6, there is adequate time for the operator toevaluate unit conditions and respond by manually startingindividual
: systems, pumps, and other equipment to mitigate theconsequences of an abnormal condition or accident.
RCSpressure and temperature are very low, and many EScomponents are administratively controlled or otherwise prevented from actuating to prevent inadvertent overpressurization of unit systems.3, 4. Reactor Building Pressure
-High and Reactor BuildingPressure
-High HighThe RB Pressure
-High and RB Pressure
-High High actuation Functions of ESPS shall be OPERABLE in MODES 1, 2, 3, and 4when the potential for a HELB exists. In MODES 5 and 6, the unitconditions are such that there is insufficient energy in the primaryand secondary systems to raise the containment pressure to eitherthe RB Pressure
-High or RB Pressure
-High High actuation setpoints.
Furthermore, in MODES 5 and 6, there is adequate timefor the operator to evaluate unit conditions and respond bymanually starting individual
: systems, pumps, and other equipment to mitigate the consequences of an abnormal condition or accident.
RCS pressure and temperature are very low and many EScomponents are administratively controlled or otherwise prevented from actuating to prevent inadvertent overpressurization of unitsystems.ACTIONS Required Actions A and B apply to all ESPS input instrumentation Parameters listed in Table 3.3.5-1.A Note has been added to the ACTIONS indicating separate Condition entry is allowed for each Parameter.
If an input channel's trip setpoint is found nonconservative with respectto the Allowable Value, or the transmitter, instrument loop, signalprocessing electronics, or ESPS input isolation
: modules, inter-channel communication modules and processor output trip devices are foundinoperable, then all affected functions provided by that input channelshould be declared inoperable and the unit must enter the Conditions forthe particular protective Parameter affected.
OCONEE UNITS 1, 2, & 3B 3.3.5-1112/10/14 1
ESPS Input Instrumentation B 3.3.5BASESACTIONS A..1(continued)
Condition A applies when one input channel becomes inoperable in oneor more Parameters.
If one ESPS input instrument channel isinoperable, placing it in a tripped condition leaves the system in a one-out-of-two condition for actuation.
Thus, if another input channel wereto fail, the ESPS instrumentation could still perform its actuation functions.
This can be accomplished two ways: (1) by placing an input logicchannel (A, B or C) in trip with the associated Manual Trip keyswitch (the input Manual Trip channel keyswitch trips all ESPS functions in thechannel),
or (2) tripping the individual input parameter functional software through the interactive Graphical Service Monitor dialogscreen. The 4 hour Completion Time is justified based on thecontinuous monitoring and signal validation being performed and issufficient time to place a Parameter in trip. If the Parameter cannot beplaced in trip, the Operator can trip the affected channel with the use ofthe Manual Trip keyswitch until such time that the individual parameter can be placed in trip.B.1, B.2.1. B.2.2. and B.2.3Condition B applies when the Required Action and associated Completion Time of Condition A are not met or when one or moreparameters have two or more inoperable input channels.
If Condition
: Bapplies, the unit must be brought to a MODE in which the LCO does notapply. To achieve this status, the unit must be brought to at leastMODE 3 within 12 hours and, for the RCS Pressure-Low Parameter, to< 1750 psig, for the RCS Pressure-Low Low Parameter, to < 900 psig,and for the RB Pressure-High Parameter and RB Pressure-High HighParameter, to MODE 5 within 36 hours. The allowed Completion Timesare reasonable, based on operating experience, to reach the requiredunit conditions from full power conditions in an orderly manner andwithout challenging unit systems.OCONEE UNITS 1, 2, & 3B 3.3.5-1212/10/14 1
ESPS Input Instrumentation B 3.3.5BASESSURVEILLANCE The ESPS Parameters listed in Table 3.3.5-1 are subject toREQUIREMENTS CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNELCALIBRATION.
SR 3.3.5.1Performance of the CHANNEL CHECK ensures that a gross failure ofinstrumentation has not occurred.
A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels.
It isbased on the assumption that input instrument channels monitoring thesame parameter should read approximately the same value. Significant deviations between the two input instrument channels could be anindication of excessive instrument drift in one of the channels or ofsomething even more serious.
CHANNEL CHECK will detect grosschannel failure; therefore, it is key in verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined, based on a combination of thechannel instrument uncertainties, including isolation, indication, andreadability.
If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has driftedoutside its limit.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.
The CHANNEL CHECKsupplements less formal, but potentially more frequent, checks ofchannel operability during normal operational use of the displaysassociated with the LCO's required channels.
The CHANNEL CHECK requirement is met automatically.
The digitalESPS provides continuous online automatic monitoring of each of theinput signals in each channel, performs signal online validation againstrequired acceptance
: criteria, and provides hardware functional validation.
If any protective channel input signal is identified to be in the failurestatus, this condition is alarmed on the Unit Statalarm and input to theplant OAC. Immediate notification of the failure status is provided to theOperations staff.OCONEE UNITS 1, 2, & 3B 3.3.5-1312/10/14 1
ESPS Input Instrumentation B 3.3.5BASESSURVEILLANCE REQUIREMENTS (continued)
SR 3.3.5.2The SR is modified by a Note indicating that it is not applicable to theReactor Building Pressure
-High High parameter.
This surveillance does not apply to the Reactor Building Pressure High High parameter because it consists of pressure switches which provide a contact statusto the system and there is no software setpoint to verify. This SRmanually retrieves the software setpoints and verifies they are correct.The proper functioning of the processor portion of the channel iscontinuously checked by automatic cyclic self monitoring.
The properfunctioning of the processor portion of the channel is continuously checked by automatic cyclic self monitoring.
Verification of fieldinstrument setpoints is not required by this surveillance.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.3.5.3This SR has been deleted.SR 3.3.5.4CHANNEL CALIBRATION is a complete check of the input instrument
: channel, including the sensor. The test verifies that the channelresponds to a measured parameter within the necessary range andaccuracy.
CHANNEL CALIBRATION leaves the channel adjusted toaccount for instrument drift to ensure that the instrument channel remainsoperational between successive tests. CHANNEL CALIBRATION assuresthat measurement errors and processor output trip device setpoint errorsare within the assumptions of the unit specific uncertainty analysis.
CHANNEL CALIBRATIONS must be performed consistent with theassumptions of the uncertainty analysis.
Since the CHANNEL FUNCTIONAL TEST is a part of the CHANNELCALIBRATION a separate SR is not required.
The digital ESPS softwareperforms a continuous online automated cross channel check, separately for each channel, and continuous online signal error detection andvalidation.
The protection system also performs continuous onlinehardware monitoring.
The CHANNEL CALIBRATION essentially validates the self monitoring function and checks for a small set of failure modes thatare undetectable by the self monitoring function.
IIOCONEE UNITS 1, 2, & 3B 3.3.5-1412/10/14 1
ESPS Input Instrumentation B 3.3.5BASESSURVEILLANCE SR 3.3.5.4 (continued)
REQUIREMENTS The digital processors shall be rebooted as part of the calibration.
Thisverifies that the software has not changed.
Signals into the system (fromthe field instrument or at the protective system cabinet) are applied duringthe channel calibration to ensure that the instrumentation is within thespecified allowance requirements.
This, in combination with ensuring thesetpoints are entered into the software correctly per SR 3.3.5.2, verifies thesetpoints are within the Allowable Values.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.REFERENCES
: 1. UFSAR, Chapter 7.2. 10 CFR 50.49.3. EDM-102, "Instrument Setpoint/Uncertainty Calculations."
: 4. UFSAR, Chapter 15.5. 10 CFR 50.36.OCONEE UNITS 1, 2, & 3B 3.3.5-1512/10/14 1
ESPS ManualB 3.3 INSTRUMENTATION B 3.3.6 Engineered Safeguards Protective System (ESPS) Manual Initiation BASESInitiation B 3.3.6BACKGROUND The ESPS manual initiation capability allows the operator to actuate ESPSFunctions from the main control room in the absence of any other initiation condition.
This ESPS manual initiation capability is provided in the eventthe operator determines that an ESPS Function is needed and has notbeen automatically actuated.
Furthermore, the ESPS manual initiation capability allows operators to rapidly initiate Engineered Safeguards (ES)Functions.
LCO 3.3.6 covers only the system level manual initiation of theseFunctions.
LCO 3.3.5, "Engineered Safeguards Protective System (ESPS)Input Instrumentation,"
and LCO 3.3.7, "Engineered Safeguards Protective System (ESPS) Automatic Actuation Output Logic Channels,"
providerequirements on the portions of the ESPS that automatically initiate theFunctions described earlier.The ESPS manual initiation Function relies on the OPERABILITY of theautomatic actuation output logic channels (LCO 3.3.7) to perform theactuation of the systems.
A manual trip push button is provided on thecontrol room console for each of the automatic actuation output logicchannels.
Operation of the push button energizes relays whose contactsperform a logical "OR" function with the automatic actuation.
The ESPS manual initiation portion of the ESPS system is defined as theinstrumentation between the control console Trip/Reset switches and therelay output (RO) relays which actuate the end devices.
Other means ofmanual initiation, such as controls for individual ES devices, may beavailable in the control room and other unit locations.
These alternative means are not required by this LCO, nor may they be credited to fulfill therequirements of this LCO.OCONEE UNITS 1, 2, & 3B 3.3.6-112/10/14 1
ESPS ManualInitiation B 3.3.6BASESBACKGROUND (continued)
A manual actuation of the ESPS actuation functions shall be capable ofbeing initiated from the main control board Trip/Reset pushbutton switches.
Individual pushbuttons are provided for High Pressure Injection and Reactor Building (RB) Non-Essential Isolation (Channels 1 and 2),Low Pressure Injection and Low Pressure Service Water Actuation (Channels 3 and 4), RB Cooling and RB Essential Isolation (Channels 5and 6), and RB Spray (Channels 7 and 8). The manual actuation isindependent of the ESPS automatic actuation signal and is capable ofactuating all channel related actuation field components regardless of anyfailures of the automatic signal. Initiation of the manual actuation portionof ESPS will also input an actuation signal to the automatic system toprovide input to the automatic system indicating that a manual actuation has occurred.
APPLICABLE SAFETY ANALYSESThe ESPS, in conjunction with the actuated equipment, provides protective functions necessary to mitigate accidents, specifically, the loss of coolantaccident and steam line break events.The ESPS manual initiation ensures that the control room operator canrapidly initiate ES Functions.
The manual initiation trip Function is requiredas a backup to automatic trip functions and allows operators to initiateESPS whenever any parameter is rapidly trending toward its trip setpoint.
The ESPS manual initiation functions satisfy Criterion 3 of 10 CFR 50.36(Ref. 1).LCOTwo ESPS manual initiation channels of each ESPS Function shall beOPERABLE whenever conditions exist that could require ES protection ofthe reactor or RB. Two OPERABLE channels ensure that no singlerandom failure will prevent system level manual initiation of any ESPSFunction.
The ESPS manual initiation Function allows the operator toinitiate protective action prior to automatic initiation or in the event theautomatic initiation does not occur.OCONEE UNITS 1, 2, & 3B 3.3.6-212/10/14 1
ESPS Manual Initiation B 3.3.6BASESLCO(continued)
The required Function is provided by two associated channels.
as indicated in the following table:Function Associated ChannelsHPI and RB Non-Essential 1 &2Isolation, Keowee Emergency Start,Load Shed and Standby BreakerInput, and Keowee Standby BusFeeder Breaker Input.LPI 33&4RB Cooling and RB Essential 5 & 6isolation RB Spray 7 & 8APPLICABILITY The ESPS manual initiation Functions shall be OPERABLE in MODES 1and 2, and in MODES 3 and 4 when the associated engineered safeguard equipment is required to be OPERABLE.
The manual initiation channelsare required because ES Functions are designed to provide protection inthese MODES. ESPS initiates systems that are either reconfigured fordecay heat removal operation or disabled while in MODES 5 and 6.Accidents in these MODES are slow to develop and would be mitigated bymanual operation of individual components.
Adequate time is available toevaluate unit conditions and to respond by manually operating the EScomponents, if required.
ACTIONSA Note has been added to the ACTIONS indicating separate Condition entry is allowed for each ESPS manual initiation Function.
A..ICondition A applies when one manual initiation channel of one or moreESPS Functions becomes inoperable.
Required Action A. I must be takento restore the channel to OPERABLE status within the next 72 hours. TheCompletion Time of 72 hours is based on operating experience andadministrative
: controls, which provide alternative means of ESPS Functioninitiation via individual component controls.
The 72 hour Completion Timeis generally consistent with the allowed outage time for the safety systemsactuated by ESPS.OCONEE UNITS 1, 2, & 3B 3.3.6-312/10/14 1
ESPS Manual Initiation B 3.3.6BASESACTIONS B.1 and B.2(continued)
With the Required Action and associated Completion Time not met, theunit must be brought to a MODE in which the LCO does not apply. Toachieve this status, the unit must be brought to at least MODE 3 within12 hours and to MODE 5 within 36 hours. The allowed Completion Timesare reasonable, based on operating experience, to reach the requiredMODES from full power conditions in an orderly manner and withoutchallenging unit systems.SURVEILLANCE SR 3.3.6.1REQUIREMENTS This SR requires the performance of a CHANNEL FUNCTIONAL TEST ofthe ESPS manual initiation.
This test verifies that the initiating circuitry isOPERABLE and will actuate the automatic actuation output logic channels.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.Failure of reactor building purge valves PR-1, 2, 3, 4, 5, 6 to close following a design basis event would cause a significant increase in the radioactive release because of the large containment leakage path introduced by these48 inch purge lines. Because of their large size, the 48 inch purge valvesare not qualified for automatic closure from their open position underaccident conditions.
Therefore, the 48 inch purge valves are maintained sealed closed (SR 3.6.3.1) in MODES 1, 2, 3, and 4 to ensure thecontainment boundary is maintained (Reference 2). Since they are sealedclosed in all modes where the Engineered Safeguards system is requiredoperable, testing of these reactor building purge valves is not required perSR 3.3.6.1.REFERENCES
: 1. 10 CFR 50.36.2. NUREG 0737, Section I1.E.4.2.6.
OCONEE UNITS 1, 2, & 3B 3.3.6-412/10/14 1
ESPS Automatic Actuation Output Logic ChannelsB 3.3.7B 3.3 INSTRUMENTATION B 3.3.7 Engineered Safeguards Protective System (ESPS) Automatic Actuation Output Logic ChannelsBASESBACKGROUND The automatic actuation output logic channels are defined as the Voters,the output relays and associated contacts.
The Voters are used to providean output signal to the output relays for the LP-1 interlock.
Since LP-1 isnot an ES valve, any inoperability of the ESPS associated with thisparticular function would require no action by TS 3.3.7. Each of thecomponents actuated by the ESPS Functions is associated with one ormore automatic actuation output logic channels.
If two-out-of-three ESPSinput instrumentation channels indicate a trip, or if channel level manualinitiation occurs, the automatic actuation output logic channel is activated and the associated equipment is actuated.
The purpose of requiring OPERABILITY of the ESPS automatic actuation output logic channels is toensure that the Functions of the ESPS can be automatically initiated in theevent of an accident.
Automatic actuation of some Functions is necessary to prevent the unit from exceeding the Emergency Core Cooling Systems(ECCS) limits in 10 CFR 50.46 (Ref. 1). It should be noted thatOPERABLE automatic actuation output logic channels alone will not ensurethat each Function can be activated; the input instrumentation channelsand actuated equipment associated with each Function must also beOPERABLE to ensure that the Functions can be automatically initiated during an accident.
LCO 3.3.7 covers only the automatic actuation output logic channels thatinitiates these Functions.
LCO 3.3.5, "Engineered Safeguards Protective System (ESPS) Input Instrumentation,"
and LCO 3.3.6, "Engineered Safeguards Protective System (ESPS) Manual Initiation,"
providerequirements on the input instrumentation and manual initiation channelsthat feed into the automatic actuation output logic channels.
The ESPS Protective Channels (computers)
A, B, and C areimplemented on two independent and redundant subsystems.
Onesubsystem, containing channels A2, B2, and C2, uses the ESPSprotective channel computers, which are installed in the ESPS cabinets.
The other sub-system, containing independent and redundant channelsAl, B1, and Cl, uses the RPS protective channel computers, which areinstalled in the RPS cabinets.
IOCONEE UNITS 1, 2, & 3B 3.3.7-112/10/14 1
ESPS Automatic Actuation Output Logic ChannelsB 3.3.7BASESBACKGROUND (continued)
Each of the independent ESPS and ESPS/RPS protective channelfunction output signals are sent to two redundant digital actuation VoterSets each comprised of an Odd and Even Voter. The Odd Voter isassociated with ESPS Automatic Actuation Output Logic Channels 1, 3, 5,and 7 while the Even Voter is associated with Channels 2, 4, 6, and 8. Oneof the Odd and Even Voter sets (Voter 2) performs the two-out-of-three voting for the actuation signals coming from the ESPS protective channels; the other independent and redundant Odd and Even Voter set (Voter 1)performs the two-out-of-three voting for the actuation signals coming fromthe ESPS/RPS sets. The independent and redundant ESPS protective safety actuation functions are duplicated in the ESPS and ESPS/RPSsubsystems.
The ESPS, in conjunction with the actuated equipment, provides protective functions necessary to mitigate accidents, specifically, the loss of coolantaccident (LOCA) and main steam line break (MSLB) events. The ESPSrelies on the OPERABILITY of the automatic actuation logic for eachcomponent to perform the actuation of the selected systems.The small break LOCA analyses assume a conservative 48 second delaytime for the actuation of High Pressure Injection (HPI) in UFSAR, Chapter15 (Ref. 2). The large break LOCA analyses assume Low PressureInjection (LPI) flow starts in 38 seconds while full LPI flow does not occuruntil 36 seconds later, or 74 seconds total (Ref. 2). This delay timeincludes allowances for Keowee Hydro Unit startup and loading, ECCSpump starts, and valve openings.
Similarly, the Reactor Building (RB)Cooling, RB Isolation, and RB Spray have been analyzed with delaysappropriate for the entire system.The ESPS automatic initiation of Engineered Safeguards (ES) Functions tomitigate accident conditions is assumed in the accident analysis and isrequired to ensure that consequences of analyzed events do not exceedthe accident analysis predictions.
Automatically actuated features includeHPI, LPI, RB Cooling, RB Spray, and RB Isolation.
OCONEE UNITS 1, 2, & 3B 3.3.7-212/10/14 1
ESPS Automatic Actuation Output Logic ChannelsB 3.3.7BASESBACKGROUND Engineered Safeguards Protective System Bypasses(continued)
There are two redundant subsystems.
The same analog input signal isfed to each subsystem.
In subsystem 1, channels Al, B1, and Clprovide the input to Voter 1 Odd and Voter 1 Even. In subsystem 2,channels A2, B2, and C2 provide input to Voter 2 Odd and Voter 2 Even.Either subsystem provides the full complement of Voters. This allows fora Manual (maintenance)
Bypass of one complete subsystem, or portionof a subsystem, without entering into an LCO Condition.
While one Voteror a set of Voters are bypassed, the ESPS function is provided by theredundant ESPS subsystem.
Placing a Voter in Manual Bypass is implemented by keyswitches locatedin the respective ESPS Actuation cabinets.
If an ESPS Voter is placed inManual Bypass, all automatic ESPS actuation functions from that specificVoter are disabled.
: However, a manual ESPS trip is still available forOperator action to initiate the ESPS safety actuation functions.
Only oneManual Bypass keyswitch for the two Odd Voters (Voter 1 Odd or Voter 2Odd) and one Manual Bypass keyswitch for the two Even Voters (Voter 1Even or Voter 2 Even) is allowed to be placed in Manual Bypass at atime. Placing an ESPS Voter in Manual Bypass is administratively controlled.
The ESPS Manual Bypass keyswitch status information issent to the Unit control room Statalarm panel and sent to the plantOperator Aid Computer (OAC).Parameter Chan-ge Enable ModeESPS Voters for subsystems 1 and 2 and Status processors can beplaced in a parameter change enable mode through the use of theParameter Change Enable keyswitches.
One keyswitch will place OddVoter 1 and the Odd Component Status processor in Parameter ChangeEnable Mode. One keyswitch will place Even 1 Voter and the EvenComponent Status processor in Parameter Change Enable Mode. OddVoter 2 and Even Voter 2 each have their own keyswitch that can beused to place each processor in Parameter Change Enable Mode.When a keyswitch is placed from the normal Operating Mode position tothe Parameter Change Enable Mode position:
* The processors continue with normal operation.
* A permissive is provided that allows the Service Unit to be used tochange the operating mode of the processors associated with thatkeyswitch.
OCONEE UNITS 1, 2, & 3B 3.3.7-312/10/14 1
ESPS Automatic Actuation Output Logic ChannelsB 3.3.7BASESBACKGROUND Parameter Change Enable Mode(continued)
With the keyswitch in the Parameter Change Enable Position thefollowing modes of operation are allowed for processors:
* Normal Operation
-with permissive for operating mode change.* Parameterization
-allows changes to specific parameters (exampleplacing a parameter into a tripped condition or performing Go/NoGotesting).
* Function Test -for disabling the application function and forcingoutput signal for testing purposes (normally not used).* Diagnostics
-for downloading new application software.
The Function Test and Diagnostics modes result in the processor ceasingits cyclic processing of the application functions.
Entry into these modesfirst requires entry into Parameterization mode and setting a separateparameter.
When a keyswitch is placed in the Parameter Change Enable ModePosition for any activity, the affected processor shall first be declared outof service.
In addition to declaring the processor out of service, whenloading or revising software in a processor, the affected ESPS voter (Set1 or Set 2) shall be placed in Bypass. Only one ESPS voter at a time isallowed to be placed into Parameter Change Enable Mode Position forsoftware loading/revision.
Each Parameter Change Enable keyswitch status information is sent tothe Statalarm panel and to the OAC via the Gateway.ESPS Parameter Change Enable keyswitches are administratively controlled (there are no hardware or software interlocks betweenchannels).
APPLICABLE Accident analyses rely on automatic ESPS actuation for protection of theSAFETY ANALYSES core and RB and for limiting off site dose levels following an accident.
Theautomatic actuation output logic is an integral part of the ESPS.The ESPS automatic actuation output logic channels satisfy Criterion 3 of10 CFR 50.36 (Ref. 3).OCONEE UNITS 1, 2, & 3B 3.3.7-412/10/14 1
ESPS Automatic Actuation Output Logic ChannelsB 3.3.7BASES (continued)
LCOThe automatic actuation output logic channels are required to beOPERABLE whenever conditions exist that could require ES protection ofthe reactor or the RB. This ensures automatic initiation of the ES requiredto mitigate the consequences of accidents.
The ESPS automatic actuation output logic channels are comprised oftwo independent and redundant subsystems.
Only one of theindependent subsystems is required to be OPERABLE.
The required Function is provided by two associated output channels asindicated in the following table:Function Associated ChannelsHPI and RB Non-Essential 1 & 2Isolation, Keowee Emergency Start,Load Shed and Standby BreakerInput, and Keowee Standby BusFeeder Breaker InputLPI 3 & 4RB Cooling and RB Essential 5 & 6isolation RB Spray 7 & 8IAPPLICABILITY The automatic actuation output logic channels shall be OPERABLE inMODES 1 and 2 and in MODES 3 and 4 when the associated engineered safeguard equipment is required to be OPERABLE, because ES Functions are designed to provide protection in these MODES. Automatic actuation in MODE 5 or 6 is not required because the systems initiated by the ESPSare either reconfigured for decay heat removal operation or disabled.
Accidents in these MODES are slow to develop and would be mitigated bymanual operation of individual components.
Adequate time is available toevaluate unit conditions and respond by manually operating the EScomponents, if required.
OCONEE UNITS 1, 2, & 3B 3.3.7-512/10/14 1
ESPS Automatic Actuation Output Logic ChannelsB 3.3.7BASESACTIONS A Note has been added to the ACTIONS indicating separate Condition entry is allowed for each ESPS automatic actuation output logic channel.A.1 and A.2When one or more automatic actuation output logic channels areinoperable, the associated component(s) can be placed in their engineered safeguard configuration.
Required Action A.1 is equivalent to theautomatic actuation output logic channel performing its safety functionahead of time.In some cases, placing the component in its engineered safeguard configuration would violate unit safety or operational considerations.
Inthese cases, the component status should not be changed, but thesupported system component must be declared inoperable.
Conditions which would preclude the placing of a component in its engineered safeguard configuration
: include, but are not limited to, violation of systemseparation, activation of fluid systems that could lead to thermal shock, orisolation of fluid systems that are normally functioning.
The Completion Time of 1 hour is based on operating experience and reflects the urgencyassociated with the inoperability of a safety system component.
Required Action A.2 requires declaring the associated components of theaffected supported systems inoperable, since the true effect of automatic actuation output logic channel failure is inoperability of the supported system. The Completion Time of 1 hour is based on operating experience and reflects the urgency associated with the inoperability of a safety systemcomponent.
A combination of Required Actions A.1 and A.2 may be usedfor different components associated with an inoperable automatic actuation output logic channel.OCONEE UNITS 1, 2, & 3B 3.3.7-612/10/14 1
ESPS Automatic Actuation Output Logic ChannelsB 3.3.7BASESSURVEILLANCE SR 3.3.7.1REQUIREMENTS This SR requires manual actuation of the output channel interposing relays (referred to as Ro relays) to demonstrate OPERABILITY of therelays. The proper functioning of the processor portion of the channel iscontinuously checked by automatic cyclic self monitoring.
Failure of reactor building purge valves PR-1, 2, 3, 4, 5, 6 to close following a design basis event would cause a significant increase in the radioactive release because of the large containment leakage path introduced by these48 inch purge lines. Because of their large size, the 48 inch purge valvesare not qualified for automatic closure from their open position underaccident conditions.
Therefore, the 48 inch purge valves are maintained sealed closed (SR 3.6.3.1) in MODES 1, 2, 3, and 4 to ensure thecontainment boundary is maintained (Reference 4). Since they are sealedclosed in all modes where the Engineered Safeguards system is requiredoperable, testing of these reactor building purge valves is not required perSR 3.3.7.1.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.3.7.2SR 3.3.7.2 is the performance of a CHANNEL FUNCTIONAL TEST. Thefunctional test consists of rebooting the digital processors.
This verifiesthat the software has not changed.Failure of reactor building purge valves PR-1, 2, 3, 4, 5, 6 to close following a design basis event would cause a significant increase in the radioactive release because of the large containment leakage path introduced by these48 inch purge lines. Because of their large size, the 48 inch purge valvesare not qualified for automatic closure from their open position underaccident conditions.
Therefore, the 48 inch purge valves are maintained sealed closed (SR 3.6.3.1) in MODES 1, 2, 3, and 4 to ensure thecontainment boundary is maintained (Reference 4). Since they are sealedclosed in all modes where the Engineered Safeguards system is requiredoperable, testing of these reactor building purge valves is not required perSR 3.3.7.2.OCONEE UNITS 1, 2, & 3B 3.3.7-712/10/14 1
ESPS Automatic Actuation Output Logic ChannelsB 3.3.7BASESSURVEILLANCE REQUIREMENTS SR 3.3.7.2 (continued)
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.The digital ESPS software performs a continuous online automated crosschannel check, separately for each channel, and continuous online signalerror detection and validation.
The protection system also performscontinual online hardware monitoring.
The CHANNEL FUNCTIONAL TEST essentially validates the self monitoring function and checks for asmall set of failure modes that are undetectable by the self monitoring function.
REFERENCES
: 1. 10 CFR 50.46.2. UFSAR, Chapter 15.3. 10 CFR 50.36.4. NUREG 0737, Section II.E.4.2.6.
OCONEE UNITS 1, 2, & 3B 3.3.7-812/10/14 1
LPSW RB Waterhammer Prevention Circuitry B 3.3.27B 3.3 INSTRUMENTATION B 3.3.27 Low Pressure Service Water (LPSW) Reactor Building (RB) Waterhammer Prevention Circuitry BASESBACKGROUND NRC Generic Letter 96-06 identified three issues of concern relative to effectsof fluid in piping following postulated design basis events. One area of concernis the cooling water system piping serving the containment air coolers.
TheLow Pressure Service Water (LPSW) system provides cooling water to thesafety related Reactor Building Cooling Units (RBCUs),
non-safety relatedReactor Building Auxiliary Cooling Units (RBACs) and non-safety relatedReactor Coolant Pump Motor (RCPM) coolers.
There is a possibility ofwaterhammer in the LPSW piping inside containment during either a Loss-of-Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) concurrent witha loss of off-site power (LOOP) without means to prevent waterhammer.
The LPSW RB Waterhammer Prevention System (WPS) is composed ofcheck valves, active pneumatic discharge isolation valves, and activecontrollable vacuum breaker valves. The LPSW RB Waterhammer Prevention Circuitry isolates LPSW to the RBCUs, RBACs and RCPM coolersany time the LPSW header pressure decreases significantly, such as during aLOOP event or LPSW pump failure during normal operations.
The isolation function prevents and/or minimizes the potential waterhammers in theassociated piping. The LPSW RB Waterhammer Prevention Circuitry willalso re-establish flow to the containment air coolers following WPS actuation once the LPSW system has repressurized.
The RBCU fans and RBCU cooling water motor operated return valves areEngineered Safeguards (ES) features.
On an ES actuation, these valvesopen. The LPSW RB Waterhammer Prevention Pneumatic Discharge Isolation Valves are designed to close on low LPSW supply header pressureand re-open when the LPSW supply header pressure is restored.
The LPSWRB Waterhammer Prevention Controllable Vacuum Breaker Valves aredesigned to open on low LPSW pressure and re-close when LPSW pressureis restored.
The LPSW RB Waterhammer Prevention Pneumatic Discharge Isolation Valves fail open on loss of instrument air. During normal operation, a controlsolenoid valve in the instrument air supply to eachOCONEE UNITS 1, 2, & 3B 3.3.27-112/10114 1
LPSW RB Waterhammer Prevention Circuitry B 3.3.27BASESBACKGROUND (continued)
LPSW RB Waterhammer Prevention Pneumatic Discharge Isolation Valve is energized to vent air from the actuator to maintain the isolation valves in the open position.
On loss of two of four of the analog input signalsfor the LPSW RB Waterhammer Prevention Isolation Circuitry, the 3-waycontrol solenoid valve is de-energized to align the air accumulator with thepneumatic operator; thereby closing the LPSW RB Waterhammer Prevention Pneumatic Discharge Isolation Valve(s).
LPSW RB Waterhammer Prevention Controllable Vacuum Breaker Valves are located downstream ofthe pneumatic discharge isolation valves. The LPSW RB Waterhammer Prevention Controllable Vacuum Breaker Valves are normally closed. Theyopen simultaneously with the closing of the LPSW RB Waterhammer Prevention Pneumatic Discharge Isolation Valves in order to break vacuum inthe return header by energizing the control solenoid valve.The LPSW RB Waterhammer Prevention Circuitry contains four analog sensorchannels and two digital actuation logic channels.
Only three analog sensorchannels are required to support OPERABILITY.
Each analog sensor channelcontains a safety grade pressure transmitter and current switch. The two digitalactuation logic channels consist of safety grade relays in a two-out-of-two logicconfiguration.
The actuation of the LPSW RB Waterhammer Prevention Circuitry requires two of the three required LPSW pressure signals suppliedfrom the LPSW header pressure transmitters.
APPLICABLE SAFETY ANALYSESIn a LOOP event, the LPSW RB Waterhammer Prevention Circuitry isolates the cooling water flow to the RBCUs, RBACs and RCPMcooler on low LPSW supply header pressure prior to LPSW pump restart toprevent waterhammers.
The LPSW RB Waterhammer Prevention Circuitry will also re-establish flow to the containment air coolers following WPSactuation once the LPSW system has repressurized.
Isolating and re-establishing the LPSW flowpath ensures that Containment Integrity andContainment Heat Removal functions are maintained.
The RBCU Fans presently have a 3 minute delay to re-start following ESactivation.
LPSW flow will be restored to the RBCUs prior to the RBCU fanrestart.
This ensures the Containment Heat Removal function is unaffected.
The LPSW RB Waterhammer Prevention Circuitry satisfies Criterion 3 of 10CFR 50.36 (Ref. 1).OCONEE UNITS 1, 2, & 3B 3.3.27-212/10/14 1
LPSW RB Waterhammer Prevention Circuitry B 3.3.27BASES (continued)
LCO Three LPSW RB Waterhammer Prevention analog channels and two digitallogic channels shall be OPERABLE.
Each analog sensor channel contains asafety related pressure transmitter and current switch. The two digital logicchannels consist of safety related relays. The LPSW RB Waterhammer Prevention Circuitry design ensures that a single active failure will not preventthe circuitry and associated components from performing the intended safetyfunctions.
There are four analog channels, but only three are required to supportOPERABILITY.
These three analog channels are configured in a two out ofthree control logic scheme that will isolate/reset the LPSW RB Waterhammer Prevention Circuitry.
The LPSW RB Waterhammer Prevention Circuitry willclose/open the four LPSW RB Pneumatic Discharge Isolation Valves whenLPSW pressure is either low or returns to normal. Either digital logic channelwill trip/restore the flow path.The actuation logic used for the LPSW RB Waterhammer Prevention Circuitry is similar to other safety related circuitry currently being used. The LCOallowed required action and Completion Times are acceptable based on thenumber of channels normally available.
Though one of the four analogchannels can be out of service for an extended period, it is not a normalpractice.
When one required analog channel is taken out of service, the two out ofthree analog control logic scheme is reduced to a two out of two analogcontrol logic scheme. This control logic scheme will trip/reset the digitalchannels on decreasing/increasing supply header pressure.
Failure of an analog channel while in the two out of two control logic modewill reduce the control logic to a one out of two control logic scheme. Thiscontrol logic is unacceptable because a failure will prevent the LPSW RBWaterhammer Prevention Circuitry from working as required.
The two digital channels are triggered by two of four analog channelsconsisting of a pressure transmitter/current switch. On decreasing/increasing supply header pressure, two of four analog channels will trip/reset the digitalchannels.
If one of the two digital channels is inoperable or out of service,the system is no longer single failure proof.OCONEE UNITS 1, 2, & 3B 3.3.27-312/10/14 1
LPSW RB Waterhammer Prevention Circuitry B 3.3.27BASES (continued)
APPLICABILITY The LPSW RB Waterhammer Prevention Circuitry is required to beOPERABLE in MODES 1, 2, 3, and 4. This ensures LPSW is available tosupport the OPERABILITY of the equipment serviced by the LPSW system.In MODES 5 and 6, the probability and consequences of the events that theLPSW System supports is reduced due to the pressure and temperature limitations of these MODES. As a result, the LPSW RB Waterhammer Prevention Circuitry is not required to be OPERABLE in MODES 5 and 6.IACTIONSA..1If one required LPSW RB Waterhammer Prevention analog channel isinoperable, the LPSW RB Waterhammer Prevention Circuitry is no longersingle failure proof and the control logic scheme is reduced to a two out oftwo configuration.
Required Action A.1 requires the LPSW RB Waterhammer Prevention analog channels be restored to OPERABLE status within 7 days.The 7 day Completion Time takes into account the allowed outage times ofsimilar systems, reasonable time for repairs, and the low probability of anevent occurring during this period.B.1If one required LPSW RB Waterhammer Prevention digital logic channel isinoperable, the LPSW RB Waterhammer Prevention Circuitry is not singlefailure proof. Required Action B.1 requires the digital channels be restored toOPERABLE status within 7 days.The 7 day Completion Time takes into account the allowed outage times ofsimilar systems, reasonable time for repairs, and the low probability of anevent occurring during this period.OCONEE UNITS 1, 2, & 3B 3.3.27-412/10/14 1
LPSW RB Waterhammer Prevention Circuitry B 3.3.27BASESACTIONS C.1 and C.2(continued)
If two or more required LPSW RB Waterhammer Prevention analogchannel(s) or two digital logic channel(s) are inoperable or the RequiredActions and associated Completion Times of Condition A or B are not met,the WPS must be configured in order to assure the Containment Integrity andHeat removal functions are maintained.
To achieve this status, actions toprevent automatic closing by manually opening (remote or local) two LPSWRB Waterhammer Prevention Pneumatic Discharge Isolation valves in thesame header shall be completed immediately and actions to repair theinoperable equipment shall be taken immediately.
LCO 3.7.7 will also applywhen the LPSW RB Waterhammer Prevention Pneumatic Discharge Isolation valves in the same header are opened.SURVEILLANCE SR 3.3.27.1REQUIREMENTS Performance of the CHANNEL CHECK ensures that a gross failure ofinstrumentation has not occurred.
A CHANNEL CHECK is normally acomparison of the parameter indicated on one channel to a similar parameter on other channels.
It is based on the assumption that analog instrument channels monitoring the same parameter should read approximately the samevalue. Significant deviations between the two analog instrument channelscould be an indication of excessive instrument drift in one of the channels or ofsomething even more serious.
CHANNEL CHECK will detect gross channelfailure; therefore, it is key in verifying that the instrumentation continues tooperate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined, based on a combination of the channelinstrument uncertainties, including isolation, indication, and readability.
If achannel is outside the criteria, it may be an indication that the transmitter or thesignal processing equipment has drifted outside its limit.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.OCONEE UNITS 1, 2, & 3B 3.3.27-512/10/14 1
LPSW RB Waterhammer Prevention Circuitry B 3.3.27BASESSURVEILLANCE SR 3.3.27.1 (continued)
REQUIREMENTS The CHANNEL CHECK supplements less formal, but potentially morefrequent, checks of channel operability during normal operational use of thedisplays associated with the LCO's required channels.
SR 3.3.27.2A CHANNEL FUNCTIONAL TEST is performed on each channel to ensure thecircuitry will perform its intended function.
The Surveillance Frequency isbased on operating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.SR 3.3.27.3A CHANNEL CALIBRATION is a complete check of the analog instrument
: channel, including the sensor. The test verifies that the channel responds toa measured parameter within the necessary range and accuracy.
TheCHANNEL CALIBRATION leaves the components adjusted to account forinstrument drift to ensure that the circuitry remains operational betweensuccessive tests. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.REFERENCES
: 1. 10 CFR 50.36.OCONEE UNITS 1, 2, & 3B 3.3.27-612/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5B 3.6 CONTAINMENT SYSTEMSB 3.6.5 Reactor Building Spray and Cooling SystemsBASESBACKGROUND The Reactor Building Spray and Reactor Building Cooling systems providecontainment atmosphere cooling to limit post accident pressure andtemperature in containment to less than the design values. Reduction ofcontainment pressure and the iodine removal capability of the sprayreduces the release of fission product radioactivity from containment to theenvironment, in the event of an accident, to within limits. The ReactorBuilding Spray and Reactor Building Cooling systems are designed to meetONS Design Criteria (Ref. 1).The Reactor Building Cooling System and Reactor Building Spray Systemare Engineered Safeguards (ES) systems.
They are designed to ensurethat the heat removal capability required during the post accident periodcan be attained.
The Reactor Building Spray System and Reactor BuildingCooling System provide containment heat removal operation.
The ReactorBuilding Spray System and Reactor Building Cooling System providemethods to limit and maintain post accident conditions to less than thecontainment design values.Reactor Building Spray SystemThe Reactor Building Spray System consists of two separate trains ofequal capacity, each capable of meeting the design basis. Each trainincludes a reactor building spray pump, spray headers,
: nozzles, valves,piping and a flow indicator.
Each train is powered from a separate ES bus.The borated water storage tank (BWST) supplies borated water to theReactor Building Spray System during the injection phase of operation.
Inthe recirculation mode of operation, Reactor Building Spray System pumpsuction is manually transferred to the reactor building sump.OCONEE UNITS 1, 2, & 3B 3.6.5-112/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESBACKGROUND Reactor Building Spray System (continued)
The Reactor Building Spray System provides a spray of relatively coldborated water into the upper regions of containment to reduce thecontainment pressure and temperature and to reduce the concentration offission products in the containment atmosphere during an accident.
In therecirculation mode of operation, heat is removed from the reactor buildingsump water by the decay heat removal coolers.
Each train of the ReactorBuilding Spray System provides adequate spray coverage to meet thesystem design requirements for containment heat removal.The Reactor Building Spray System is actuated automatically by acontainment High-High pressure signal. An automatic actuation opens theReactor Building Spray System pump discharge valves and starts the twoReactor Building Spray System pumps.Reactor Buildingq Cooling SystemThe Reactor Building Cooling System consists of three reactor buildingcooling trains. Each cooling train is equipped with cooling coils, and anaxial vane flow fan driven by a two speed electric motor.During normal unit operation, typically two reactor building cooling trainswith two fans operating at low speed or high speed, serve to cool thecontainment atmosphere.
Low speed cooling fan operation is available during periods of lower containment heat load. The third unit is usually onstandby.
Upon receipt of an emergency signal, the operating cooling fansrunning at low speed or high speed will automatically trip, then restart in lowspeed after a 3 minute delay, and any idle unit is energized in low speedafter a 3 minute delay. The fans are operated at the lower speed duringaccident conditions to prevent motor overload from the higher densityatmosphere.
The common LPSW return header will split into two new headersdownstream of the Reactor Building Cooling Units (RBCUs).
Eachheader will contain two pneumatic discharge isolation valves and will becapable of full LPSW flow. The headers will be rejoined downstream ofthe discharge isolation valves into a common return.APPLICABLE The Reactor Building Spray System and Reactor Building Cooling SystemSAFETY ANALYSES reduce the temperature and pressure following an accident.
The limitingaccidents considered are the loss of coolant accident (LOCA) and thesteam line break. The postulated accidents are analyzed, with regard tocontainment ES systems, assuming the loss of one ES bus. This is theOCONEE UNITS 1, 2, & 3B 3.6.5-212/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESAPPLICABLE worst-case single active failure, resulting in one train of the Reactor BuildingSAFETY ANALYSES Spray System and one train of the Reactor Building Cooling System being(continued) inoperable.
The analysis and evaluation show that, under the worst-case scenario (LOCAwith worst-case single active failure),
the highest peak containment pressureis 57.75 psig. The analysis shows that the peak containment temperature is283.1 OF. Both results are less than the design values. The analyses andevaluations assume a power level of 2619 MWt, one reactor building spraytrain and two reactor building cooling trains operating, and initial (pre-accident) conditions of 80&deg;F and 15.9 psia. The analyses also assume a delayedinitiation to provide conservative peak calculated containment pressure andtemperature responses.
The Reactor Building Spray System total delay time of approximately 142 seconds includes Keowee Hydro Unit startup (for loss of offsite power),reactor building spray pump startup, and spray line filling (Ref. 2).Reactor building cooling train performance for post accident conditions isgiven in Reference
: 2. The result of the analysis is that any combination of twotrains can provide 100% of the required cooling capacity during the postaccident condition.
The train post accident cooling capacity under varyingcontainment ambient conditions is also shown in Reference 2.Reactor Building Cooling System total delay time of 3 minutes includes KHUstartup (for loss of offsite power) and allows all ES equipment to start beforethe Reactor Building Cooling Unit on the associated train is started.
Thisimproves voltages at the 600V and 208V levels for starting loads (Ref. 2).The Reactor Building Spray System and the Reactor Building Cooling Systemsatisfy Criterion 3 of 10 CFR 50.36 (Ref. 3).LCO During an accident, a minimum of two reactor building cooling trains and onereactor building spray train are required to maintain the containment pressureand temperature following a LOCA. Additionally, one reactor building spraytrain is required to remove iodine from the containment atmosphere andmaintain concentrations below those assumed in the safety analysis.
Toensure that these requirements are met, two reactor building spray trains andthree reactor building cooling trains must be OPERABLE in MODES 1 and 2.In MODES 3 or 4, one reactor building spray train and two reactor buildingcooling trains are required to be OPERABLE.
The LCO is provided with anote that clarifies this requirement.
Therefore, in the event of an accident, theminimum requirements are met, assuming the worst-case single active failureoccurs.OCONEE UNITS 1, 2, & 3B 3.6.5-312/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESLCO(continued)
Each reactor building spray train shall include a spray pump, sprayheaders,
: nozzles, valves, piping, instruments, and controls to ensure anOPERABLE flow path capable of taking suction from the BWST (via theLPI System) upon an Engineered Safeguards Protective System signal andmanually transferring suction to the reactor building sump. TheOPERABILITY of RBS train flow instrumentation is not required forOPERABILITY of the corresponding RBS train because system resistance hydraulically maintains adequate NPSH to the RBS pumps and manualthrottling of RBS flow is not required.
During an event, LPI train flow mustbe monitored and controlled to support the RBS train pumps to ensure thatthe NPSH requirements for the RBS pumps are not exceeded.
If the flowinstrumentation or the capability to control the flow in a LPI train isunavailable then the associated RBS train's OPERABILITY is affected untilsuch time as the LPI train is restored or the associated LPI pump is placedin a secured state to prevent actuation during an event.Each reactor building cooling train shall include cooling coils, fusibledropout plates or duct openings, an axial vane flow fan, instruments, valves, and controls to ensure an OPERABLE flow path. Two headers ofthe LPSW RB Waterhammer Prevention Discharge Isolation Valves arerequired to support flowpath OPERABILITY or one header of LPSW RBWaterhammer Prevention Discharge Isolation Valves shall be manuallyopened (remote or local) to prevent automatic closure.
Valve LPSW-108shall be locked open to support system OPERABILITY.
IAPPLICABILITY In MODES 1, 2, 3, and 4, an accident could cause a release of radioactive material to containment and an increase in containment pressure andtemperature, requiring the operation of the reactor building spray trains andreactor building cooling trains.In MODES 5 and 6, the probability and consequences of these events arereduced due to the pressure and temperature limitations of these MODES.Thus, the Reactor Building Spray System and the Reactor Building CoolingSystem are not required to be OPERABLE in MODES 5 and 6.ACTIONS The Actions are modified by a Note indicating that the provisions ofLCO 3.0.4 do not apply for Unit 2 only. As a result, this allows entry into aMODE or other specified condition in the Applicability with the LCO not metafter performance of a risk assessment addressing inoperable systems andcomponents, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, andOCONEE UNITS 1, 2, & 3B 3.6.5-412/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESACTIONS(continued) establishment of risk management
: actions, if appropriate.
The riskassessment may use quantitative, qualitative, or blended approaches and the risk assessment will be conducted using the plant program,procedures, and criteria in place to implement 10 CFR 50.65(a)(4),
whichrequires that risk impacts of maintenance activities to be assessed andmanaged.
The risk assessment must take into account all inoperable Technical Specifications equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. Therisk assessments will be conducted using the procedures and guidanceendorsed by Regulatory Guide 1.182, "Assessing and Managing RiskBefore Maintenance Activities at Nuclear Power Plants."
Regulatory Guide 1.1 82 endorses the guidance in Section 11 of NUMARC 93-01,"Industry Guideline for Monitoring the Effectiveness of Maintenance atNuclear Power Plants."
These documents address general guidance forconduct of the risk assessment, quantitative and qualitative guidelines forestablishing risk management
: actions, and example risk management actions.
These include actions to plan and conduct other activities in amanner that controls overall risk, increased risk awareness by shift andmanagement personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment ofbackup success paths or compensatory measures),
and determination that the proposed MODE change is acceptable.
Consideration shouldalso be given to the probability of completing restoration such that therequirements of the LCO would be met prior to the expiration ofACTIONS Completion Times that would require exiting the Applicability.
The risk assessment does not have to be documented.
There is a small subset of systems and components that have beendetermined (Ref: B&W owners group generic qualitative riskassessments-attachment to TSTF-359, Rev. 9, "B&W owners groupQualitative Risk Assessment for Increased Flexibility in MODERestraints,"
Framatome Technologies BAW-2383, October 2001.) to beof higher risk significance for which an LCO 3.0.4 exemption would not beallowed.
For Oconee these are the Decay Heat Removal System (DHR)entering MODES, 5 and 4; Keowee Hydro Units entering MODES 1-5;and the emergency feedwater system (EFW) entering MODE 1. TheReactor Spray and Cooling System is not one of the higher risksignificant systems noted.The provisions of this Note should not be interpreted as endorsing thefailure to exercise the good practice of restoring systems or components toOPERABLE status before entering an associated MODE or other specified Condition in the Applicability.
OCONEE UNITS 1, 2, & 3B 3.6.5-512/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESACTIONS A.1(continued)
With one reactor building spray train inoperable in MODE 1 or 2, theinoperable reactor building spray train must be restored to OPERABLEstatus within 7 days. In this Condition, the remaining OPERABLE sprayand cooling trains are adequate to perform the iodine removal andcontainment cooling functions.
The 7 day Completion Time takes intoaccount the redundant heat removal capability afforded by the OPERABLEreactor building spray train, reasonable time for repairs, and the lowprobability of an accident occurring during this period.The 14 day portion of the Completion Time for Required Action A.1 isbased upon engineering judgment.
It takes into account the low probability of coincident entry into two Conditions in this LCO coupled with the lowprobability of an accident occurring during this time. Refer to Section 1.3,Completion Times, for a more detailed discussion of the purpose of the"from discovery of failure to meet the LCO" portion of the Completion Time.B. 1With one of the reactor building cooling trains inoperable in MODE 1 or 2,the inoperable reactor building cooling train must be restored toOPERABLE status within 7 days. The components in this degradedcondition provide iodine removal capabilities and are capable of providing at least 100% of the heat removal needs after an accident.
The 7 dayCompletion Time was developed taking into account the redundant heatremoval capabilities afforded by combinations of the Reactor BuildingSpray System and Reactor Building Cooling System and the low probability of an accident occurring during this period.The 14 day portion of the Completion Time for Required Action B. I isbased upon engineering judgment.
It takes into account the low probability of coincident entry into two Conditions in this LCO coupled with the lowprobability of an accident occurring during this time. Refer to Section 1.3for a more detailed discussion of the purpose of the "from discovery offailure to meet the LCO" portion of the Completion Time.C.1With one reactor building spray train and one reactor building cooling traininoperable in MODE 1 or 2, at least one of the inoperable trains must berestored to OPERABLE status within 24 hours. In this Condition, theremaining OPERABLE spray and cooling trains are adequate to provideiodine removal capabilities and are capable of providing at least 100% ofOCONEE UNITS 1, 2, & 3B 3.6.5-612/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESACTIONS C.1 (continued) the heat removal needs after an accident.
The 24 hour Completion Timetakes into account the heat removal capability afforded by the remaining OPERABLE spray train and cooling trains, reasonable time for repairs, andthe low probability of an accident occurring during this period.D.1If the Required Action and associated Completion Time of Condition A, Bor C are not met, the unit must be brought to a MODE in which the LCO, asmodified by the Note, does not apply. To achieve this status, the unit mustbe brought to at least MODE 3 within 12 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the requiredunit conditions from full power conditions in an orderly manner and withoutchallenging unit systems.E.1With one of the required reactor building cooling trains inoperable in MODE3 or 4, the required reactor building cooling train must be restored toOPERABLE status within 24 hours.The 24 hour Completion Time is reasonable based on engineering judgement taking into account the iodine and heat removal capabilities ofthe remaining required train of reactor building spray and cooling.F._1With one required reactor building spray train inoperable in MODE 3 or 4,the required reactor building spray train must be restored to OPERABLEstatus within 24 hours. The 24 hour Completion Time is reasonable basedon engineering judgement taking into account the heat removal capabilities of the remaining required trains of reactor building cooling.G. 1If the Required Actions and associated Completion Times of Condition Eor F of this LCO are not met, the unit must be brought to a MODE in whichthe LCO does not apply. To achieve this status, the unit must be broughtto MODE 5 within 36 hours. The allowed Completion Times arereasonable, based on operating experience, to reach the required unitOCONEE UNITS 1, 2, & 3B 3.6.5-712/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESACTIONS G.1 (continued) conditions from full power conditions in an orderly manner and withoutchallenging unit systems.H..1With two reactor building spray trains, two reactor building cooling trains orany combination of three or more reactor building spray and reactorbuilding cooling trains inoperable in MODE 1 or 2, the unit is in a condition outside the accident analysis.
Therefore, LCO 3.0.3 must be enteredimmediately.
With any combination of two or more required reactor building spray andreactor building cooling trains inoperable in MODE 3 or 4, the unit is in acondition outside the accident analysis.
Therefore, LCO 3.0.3 must beentered immediately.
SURVEILLANCE SR 3.6.5.1REQUIREMENTS Verifying the correct alignment for manual and non-automatic poweroperated valves in the reactor building spray and cooling flow path providesassurance that the proper flow paths will exist for Reactor Building Sprayand Cooling System operation.
This SR does not apply to valves that arelocked, sealed, or otherwise secured in position, since these were verifiedto be in the correct position prior to locking,
: sealing, or securing.
Similarly, this SR does not apply to automatic valves since automatic valves actuateto their required position upon an accident signal. This SR also does notapply to valves that cannot be inadvertently misaligned, such as checkvalves. This SR does not require any testing or valve manipulation.
Rather, it involves verification, through a system walkdown, that thosevalves outside containment and capable of potentially being mispositioned are in the correct position.
The Surveillance Frequency is based onoperating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.OCONEE UNITS 1, 2, & 3B 3.6.5-812/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESSURVEILLANCE REQUIREMENTS (continued)
SR 3.6.5.2Operating each required reactor building cooling train fan unit for>_ 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly.
It also ensures that blockage, fan ormotor failure, or excessive vibration can be detected for corrective action.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.6.5.3Verifying that each required Reactor Building Spray pump's developed head at the flow test point is greater than or equal to the requireddeveloped head ensures that spray pump performance has not degradedduring the cycle. Flow and differential pressure are normal tests ofcentrifugal pump performance required by Section XI of the ASME Code(Ref. 4). Since the Reactor Building Spray System pumps cannot betested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and isindicative of overall performance.
Such inservice tests confirm component OPERABILITY, trend performance, and may detect incipient failures byindicating abnormal performance.
The Frequency of this SR is inaccordance with the Inservice Testing Program.SR 3.6.5.4Verifying the containment heat removal capability provides assurance thatthe containment heat removal systems are capable of maintaining containment temperature below design limits following an accident.
Thistest verifies the heat removal capability of the Low Pressure Injection (LPI)Coolers and Reactor Building Cooling Units. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk andis controlled under the Surveillance Frequency Control Program.OCONEE UNITS 1, 2, & 3B 3.6.5-912/10/14 1
Reactor Building Spray and Cooling SystemsB 3.6.5BASESSURVEILLANCE REQUIREMENTS (continued)
SR 3.6.5.5 and 3.6.5.6These SRs require verification that each automatic reactor building sprayand cooling valve actuates to its correct position and that each reactorbuilding spray pump starts upon receipt of an actual or simulated actuation signal. The test will be considered satisfactory if visual observation andcontrol board indication verifies that all components have responded to theactuation signal properly; the appropriate pump breakers have closed, andall valves have completed their travel. This SR is not required for valvesthat are locked, sealed, or otherwise secured in position underadministrative controls.
The Surveillance Frequency is based onoperating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.SR 3.6.5.7This SR requires verification that each required reactor building coolingtrain actuates upon receipt of an actual or simulated actuation signal. Thetest will be considered satisfactory if control board indication verifies that allcomponents have responded to the actuation signal properly, theappropriate valves have completed their travel, and fans are running at halfspeed. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.6.5.8With the reactor building spray header isolated and drained of anysolution, station compressed air is introduced into the spray headers.
ThisSR requires verification that each spray nozzle is unobstructed following activities which could cause nozzle blockage.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk andis controlled under the Surveillance Frequency Control Program.REFERENCES
: 1. UFSAR, Section 3.1.2. UFSAR, Section 6.2.3. 10 CFR 50.36.4. ASME, Boiler and Pressure Vessel Code, Section Xl.OCONEE UNITS 1, 2, & 3B 3.6.5-1012/10/14 1
LPSW SystemB 3.7.7B 3.7 PLANT SYSTEMSB 3.7.7 Low Pressure Service Water (LPSW) SystemBASESBACKGROUND The LPSW System provides a heat sink for the removal of process andoperating heat from safety related components during a transient oraccident.
During normal operation and normal shutdown, the LPSWSystem also provides this function for various safety related andnonsafety related components.
The LPSW system for Unit 1 and Unit 2 is shared and consists of threeLPSW pumps which can supply multiple combinations of path ways tosupply required components.
The LPSW system for Unit 3 consists oftwo LPSW pumps which can supply multiple combinations of path waysto supply required components.
Although multiple combinations of pathways exist, only one flow path is necessary, since no single failure of anactive component can prevent the LPSW system from supplying necessary components.
The pumps and valves are remote manuallyaligned, except in the unlikely event of a loss of coolant accident (LOCA)or other accidents.
The pumps are automatically started upon receipt ofan Engineered Safeguards actuation signal, and automatic valves arealigned to their post accident positions.
The LPSW System also providescooling directly to the Reactor Building Cooling Units (RBCU) and LowPressure Injection
: coolers, turbine driven EFW pump, HPI pump motorcoolers, and the motor driven EFW pumps.GL 96-06 required consideration of waterhammer inside containment during a LOCA or MSLB combined with a loss of offsite power (LOOP)event. As a result, the LPSW Reactor Building (RB) Waterhammer Prevention System (WPS) was added to maintain LPSW piping watersolid inside containment during any event that causes a loss of LPSWsystem pressure.
The WPS is fully automatic.
Other functions of theWPS are addressed by LCO 3.3.27 and LCO 3.6.5.Additional information about the design and operation of the LPSWSystem, along with a list of the components served, is presented in theUFSAR, Section 9.2.2 (Ref. 1).APPLICABLE The primary safety function of the LPSW System is, in conjunction with aSAFETY ANALYSES 100% capacity reactor building cooling system, (a combination of thereactor building spray and reactor building air coolers) to remove coredecay heat following a design basis LOCA, as discussed in the UFSAR,OCONEE UNITS 1, 2, & 3B 3.7.7-112/10/14 1
LPSW SystemB 3.7.7BASESAPPLICABLE Section 6.3 (Ref. 2). This provides for a gradual reduction in theSAFETY ANALYSES temperature of the fluid, as it is supplied to the Reactor Coolant System(continued)
(RCS) by the High Pressure and Low Pressure Injection pumps.The LPSW System is designed to perform its function with a single activefailure of any component, assuming loss of offsite power.The LPSW System also cools the unit from Decay Heat Removal (DHR)System entry conditions, to MODE 5 during normal and post accidentoperation.
The time required for this evolution is a function of the numberof DHR System trains that are operating.
One LPSW pump per unit anda flowpath is sufficient to remove decay heat during subsequent operations in MODES 5 and 6. This assumes a maximum LPSW Systemtemperature of 90&deg;F occurring simultaneously with maximum heat loadson the system.The LPSW System satisfies Criterion 3 of 10 CFR 50.36 (Ref. 2).LCO For the LPSW system shared by Units 1 and 2, three LPSW pumps arerequired to be OPERABLE to provide the required redundancy to ensurethat the system functions to remove post accident heat loads, assumingthe worst case single active failure occurs coincident with the loss ofoffsite power. The LCO is modified by a Note which requires only twoLPSW pumps to be OPERABLE for Units I or 2 if either Unit is defueledand one LPSW pump is capable of mitigating the DBA on the fueled Unit.The Units 1 and 2 LPSW System requires only two pumps to meet thesingle failure criterion provided that one of the units has been defueledand the following LPSW System loads on the defueled unit are isolated:
Reactor Building Cooling Units (RBCU), Reactor Building Auxiliary
: Coolers, Component
: Cooling, Main Turbine Oil Tank, Reactor Coolant(RC) Pumps, and Low Pressure Injection (LPI) Coolers.For the LPSW system for Unit 3, two LPSW pumps are required to beOPERABLE to provide the required redundancy to ensure that thesystem functions to remove post accident heat loads, assuming the worstcase single active failure occurs coincident with the loss of offsite power.An LPSW flow path is considered OPERABLE when the associated piping, valves, heat exchangers, and instrumentation and controlsrequired to perform the safety related function are OPERABLE.
Anycombination of pathways to supply the required components isacceptable, provided there is no single active failure which can preventsupplying necessary loads and applicable design criteria (e.g., seismicqualification) are satisfied.
OCONEE UNITS 1, 2, & 3B 3.7.7-212/10/14 1
LPSW SystemB 3.7.7BASESLCO(continued)
The LPSW WPS is considered OPERABLE when the associated leakageaccumulator, relief valves, seat leakage limits for check valves andpneumatic discharge isolation valves, closure capability of pneumatic discharge isolation valves, and opening capability of the controllable vacuum breaker valves are OPERABLE.
APPLICABILITY In MODES 1, 2, 3, and 4, the LPSW System is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the LPSW System. Therefore, the LPSW System is requiredto be OPERABLE in these MODES.In MODES 5 and 6, the OPERABILITY requirements of the LPSWSystem are determined by the systems it supports.
ACTIONSA.1If one required LPSW pump is inoperable, action must be taken torestore the required LPSW pump to OPERABLE status within 72 hours.In this Condition, the remaining OPERABLE LPSW pump(s) areadequate to perform the heat removal function.
: However, the overallreliability is reduced because a single failure in the OPERABLE LPSWpump(s) could result in loss of LPSW system function.
The 72 hourCompletion Time is based on the redundant capabilities afforded by theOPERABLE pump, and the low probability of a DBA occurring during thisperiod.B. 1If the LPSW WPS is inoperable, action shall be taken to restore therequired LPSW WPS components to OPERABLE status within 7 days.The 7 day Completion Time is based on similar systems and is considered reasonable based on engineering judgment and the low probability of aDBA occurring during the period of maintenance.
C.1 and C.2If the LPSW pump or WPS cannot be restored to OPERABLE statuswithin the associated Completion Time, the unit must be placed in aMODE in which the LCO does not apply. To achieve this status, the unitIOCONEE UNITS 1, 2, & 3B 3.7.7-312/10/14 1
LPSW SystemB 3.7.7BASESACTIONS C.1 and C.2 (continued) must be placed in at least MODE 3 within 12 hours, and in MODE 5within 60 hours.The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full powerconditions in an orderly manner and without challenging unit systems.The extended interval to reach MODE 5 provides additional time torestore the required LPSW pump and is reasonable considering that thepotential for an accident or transient is reduced in MODE 3.SURVEILLANCE SR 3.7.7.1REQUIREMENTS Verifying the correct level in the leakage accumulator will provideassurance that in the event of boundary valve leakage during a LOOPevent, there is sufficient water to keep the LPSW piping filled. Therequired water level is between half full and full, which corresponds to alevel indication of 20.5" to 41". Any level glass reading is bounded by20.5" to 41" level indication, therefore any level glass reading isconsidered acceptable.
During LPSW testing, accumulator level > 41" isacceptable because the mass of air in the accumulator is unchanged inthe short term; therefore the accumulator is still capable of performing itssafety function.
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.SR 3.7.7.2Verifying the correct alignment for manual, and power operated valves inthe LPSW System flow path provides assurance that the proper flowpaths exist for LPSW System operation.
This SR does not apply tovalves that are locked, sealed, or otherwise secured in position, sincethey are verified to be in the correct position prior to locking,
: sealing, orsecuring.
This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
This SR also does notapply to valves that cannot be inadvertently misaligned, such as checkvalves.OCONEE UNITS 1, 2, & 3B 3.7.7-412/10/14 1
LPSW SystemB 3.7.7BASESSURVEILLANCE SR 3.7.7.2 (continued)
REQUIREMENTS The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.This SR is modified by a Note indicating that the isolation of components or systems supported by the LPSW System does not affect theOPERABILITY of the LPSW System.SR 3.7.7.3The SR verifies proper automatic operation of the LPSW System valves.The LPSW System is a normally operating system that cannot be fullyactuated as part of the normal testing.
This SR is not required for valvesthat are locked, sealed, or otherwise secured in position underadministrative controls.
The Surveillance Frequency is based onoperating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.SR 3.7.7.4The SR verifies proper automatic operation of the LPSW System pumpson an actual or simulated actuation signal. The LPSW System is anormally operating system that cannot be fully actuated as part of normaltesting during normal operation.
The Surveillance Frequency is based onoperating experience, equipment reliability, and plant risk and iscontrolled under the Surveillance Frequency Control Program.SR 3.7.7.5The SR verifies proper operation of the LPSWRB Waterhammer Prevention System leakage accumulator.
Verifying adequate flow fromthe accumulator will provide assurance that in the event of boundaryvalve leakage during a LOOP event, there is sufficient water to keepLPSW piping filled.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.OCONEE UNITS 1, 2, & 3B 3.7.7-512/10/14 1
LPSW SystemB 3.7.7BASESSURVEILLANCE REQUIREMENTS (continued)
SR 3.7.7.6The SR verifies that LPSW WPS boundary valve leakage is < 20 gpm.Verifying boundary valve leakage is within limits will ensure that in theevent of a LOOP, a waterhammer will not occur, because the LPSWleakage accumulator will be able to maintain the LPSW piping watersolid.The LPSW Leakage Accumulator is designed to allow up to 25 gpm ofaggregate leakage for one minute. The boundary valve leakage is limitedto 20 gpm in order to allow five (5) gpm of miscellaneous leakage.The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under theSurveillance Frequency Control Program.IREFERENCES
: 1. UFSAR, Section 9.2.2.2. UFSAR, Section 6.3.3. 10 CFR 50.36.OCONEE UNITS 1, 2, & 3B 3.7.7-612/10/14 1}}

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