ML13316A009

Columbia Generating Station, License Amendment Request for Change to Emergency Core Cooling Systems Surveillance Requirements

ML13316A009
Person / Time
Site:	Columbia
Issue date:	10/31/2013
From:	Javorik A L Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML13316A003	List: GO2-13-151, NEDO-33813, Rev. 2, Technical Specification Change Support for Rhr/Lpci and LPCS Flow Rate. Long-Term LOCA Containment Response and ECCS/Non-LOCA Evaluations. ML13316A009
References
GO2-13-151
Download: ML13316A009 (27)

Text

Al.ex L. Javorik" .... " ; : " ": ' ;Columbia Genoratfrng StationW /u P.O. Box 968, PE04L~J Richiarid, WA 993-20968 Ph.

F. 509.377.4150 Proprietary

-Withhold under 10 CFR 2.390. Enclosure 2 contains PROPRIETARY information.

October 31, 2013G02-13-151 10 CFR 50.90U.S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555-0001 Subject COLUMBIA GENERATING

STATION, DOCKET NO. 50"397LUCENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Energy Northwest hereby requests a license amendment torevise the Columbia Generating Station Technical Specification Surveillance Requirements 3.5.1.4 and 3.5.2.5 for the Low Pressure Core Spray (LPCS) and LowPressure Coolant Injection (LPCI) pump flows. This amendment is requested toincrease pump operating margin and facilitate pump maintenance and repair.Enclosure I contains an evaluation of the proposed changes.

Attachments to Enclosure I include the following:

1. Proposed Columbia Technical Specification Changes (Mark-Up)

2. Proposed Columbia Technical Specification Changes (Re-Typed)

Enclosure 2 to this amendment request contains NEDC-33813P, "Technical Specification Change Support for RHR/LPCI and LPCS Flow Rate Long-Term LOCAContainment Response and ECCS/Non-LOCA Evaluations".

GE Hitachi NuclearEnergy (GEH) considers certain information contained in Enclosure 2 to be proprietary and, therefore, requests that it be withheld from public disclosure in accordance with 10CFR 2.390. A non-proprietary version of this document is provided in Enclosure 3.Enclosure 2 also contains the associated affidavit within the first few pages of thedocument, for the request to be withheld from public disclosure.

This letter and its enclosures contain no regulatory commitments.

Approval of the proposed amendment Is requested within one year of the date of thesubmittal.

Once approved, the amendment shall be implemented within 60 days.When Enclosure 2 Is removed from this letter, the letter and remaining Enclosures are NON-PROPRIETARY.

DD LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Page 2 of 2In accordance with 10 CFR 50.91, Energy Northwest is notifying the State ofWashington of this amendment request by transmitting a copy of this letter andenclosures to the designated State Official.

If there are any questions or if additional information is needed, please contact Ms. L. L.Williams, Licensing Supervisor, at 509-377-8148.

I declare under penalty of perjury that the foregoing is true and correct.

Executed onthe date of this letter.Respectfully, A. L. JavonkVice President, Engineering

Enclosures:

As statedcc: NRC RIV Regional Administrator NRC NRR Project ManagerNRC Senior Resident Inspector/988C AJ Rapacz -BPAI1 399 (email)JO Luce -ESFEC t RR Cowley -WDOH (email)

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 1 of 19Evaluation of Proposed Change1.0 SUMMARY DESCRIPTION This evaluation supports a License Amendment Request (LAR) to lower the flow ratesspecified in Columbia Generating Station (Columbia)

Technical Specification (TS) 3.5.1and 3.5.2 for Low Pressure Core Spray (LPCS) and Low Pressure Coolant Injection (LPCI). The TS Surveillance Requirement (SR) flow rates will be decreased from 6,350gpm to 6,200 gpm for LPCS and from 7,450 gpm to 7,200 gpm for RHR/LPCI at thesame specified differential pressure between the reactor pressure vessel (RPV) andsuppression pool (128 psid for LPCS and 26 psid for RHR/LPCI).

Implementation of this LAR will result in no physical modification to the plant. Thisproposed change has no adverse effect on the plant or plant safety.2.0 DETAILED DESCRIPTION

2.1 Background

2.1.1 LPCSThe LPCS system is one of the systems in the Emergency Core Cooling System(ECCS) network and is dedicated to assure that postulated loss of coolant accident(LOCA) consequences can be mitigated.

The LPCS system delivers water over thecore at low reactor pressures.

The primary purpose of LPCS is to provide inventory makeup and spray cooling during large breaks, which uncover the core. When assistedby the Automatic Depressurization System (ADS), LPCS also provides protection forsmall breaks.The LPCS system consists of a single motor-driven centrifugal pump, a spray sparger inthe reactor vessel above the core, piping and valves to convey water from thesuppression pool to the sparger, and associated controls and instrumentation.

LowPressure Core Spray is associated with Division 1.2.1.2 LPCIThe LPCI mode is an operating mode of the Residual Heat Removal (RHR) system andis one of the systems in the ECCS network dedicated to assure that postulated LOCAconsequences can be mitigated.

The LPCI mode delivers water to the core at lowreactor pressures.

The primary purpose of LPCI is to provide inventory makeupfollowing large pipe breaks. When assisted by ADS, LPCI also provides protection forsmall breaks.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 2 of 19The RHR system is comprised of three independent loops. Each loop contains its ownmotor-driven pump, piping, valves, instrumentation, and controls.

For the LPCI mode ofRHR, the three pumps deliver water from the suppression pool to the bypass regioninside the shroud through three separate reactor vessel penetrations and cool the coreby flooding.

The low water level or high drywell pressure

signals, which automatically initiate the LPCI mode, are also used to isolate all other modes of RHR operation andrevert system valves to the LPCI lineup. The RHR system continues in the LPCI modeuntil the operator determines that another mode of operation is needed (such ascontainment cooling) and takes action to manually initiate that mode. LPCI will not bediverted to any other mode of operation until adequate core cooling is ensured.

Nooperator actions are needed during the short term.The three RHR pumps are annotated as RHR-P-2A, RHR-P-2B, and RHR-P-2C.

RHR-P-2A is associated with Division 1 while RHR-P-2B and RHR-P-2C are associated withDivision 2.2.1.3 RHR Containment Heat RemovalThe RHR containment heat removal function is accomplished by the use of anoperational mode of the RHR system. The purpose of this system is to preventexcessive containment temperatures and pressures, thus maintaining containment integrity following a LOCA.Two of the RHR trains (A and B) are equipped with heat exchangers to provide heatremoval capability.

The RHR system's suppression pool cooling (SPC) andcontainment spray cooling (CSC) modes provide heat removal from the suppression pool and containment by pumping suppression pool water through the system's heatexchangers and discharging the water either directly back to the suppression pool (i.e.,in the SPC mode) or discharging the water to the wetwell and/or drywell spray spargers(i.e., in the CSC mode) where the water is then returned, by drainage, back to thesuppression pool. The drywell spray function also removes radioactive fission productsfrom the containment atmosphere during a LOCA. Water from the Standby ServiceWater (SW) system is pumped through the heat exchanger tube side to remove heatfrom the process water.There are no signals which automatically initiate containment cooling;

however, the SWsystem is automatically initiated by the same signals which start up the ECCS. To startRHR containment cooling after a LOCA resulting from a large break, the operatorverifies that the normally open RHR heat exchanger isolation valves are open and thenshuts the heat exchanger bypass valve. The rated containment cooling flow, 7,450gpm, can be achieved through the LPCI line, the drywell spray line, or through the testline and wetwell spray line, which directs the heat exchanger discharge directly into thesuppression pool. Thus, the design allows containment cooling simultaneously withcore flooding or containment spray. If the break size is small enough to limit reactor LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 3 of 19depressurization, the rated containment cooling flow cannot be established through theLPCI line. The operator must then direct the RHR containment cooling flow through thedrywell spray line or through the test line; however, the operator will not divert LPCI flowaway from the reactor until adequate core cooling is ensured.

When directed byprocedures, the operator may start drywell spray by shutting the LPCI injection valveand then opening the drywell spray valves. Similarly, the operator may divert the flowdirectly to the suppression pool by shutting the LPCI injection valve and then openingthe test line valve.2.1.4 RHR Shutdown CoolingThe RHR system's normal shutdown cooling mode removes reactor core decay andsensible heat from the primary reactor system to permit refueling and servicing.

Thisheat removal function is initiated manually after the reactor pressure has been reducedto less than 48 psig (2950F) by discharge of steam to the main condenser.

The RHR system's alternate shutdown cooling mode is utilized during normal plantoperation and design basis events when the normal shutdown cooling mode is notavailable to remove reactor core decay and sensible heat. This heat removal function issafety related, initiated manually and pumps suppression pool water into the core andallows the water to return to the suppression pool through the Safety/Relief Valves(SRVs).During normal plant shutdown, when the reactor vessel head has been removed, theRHR system is also designed to be capable of being aligned to assist the Fuel PoolCooling and Cleanup (FPC) system in maintaining the fuel pool temperature withinacceptable limits. In this mode the system is designed to cool water drawn from the fuelpool by passing it through an RHR system heat exchanger and then discharge thewater back to the fuel pool.2.2 Circumstances Necessitating the ChangeThis LAR requests a TS change to redefine the operating margin for safety relatedLPCS and RHR/LPCI pumps.Historically, the plant has had little operating margin with these pumps. See Figure 1 fora representation of margins.

The RHR pumps were tested prior to initial plant startupusing the actual injection flow path. RHR-P-2B produced 7,500 gpm at 28 psid on9/24/1983.

From initial installation, this pump had only 2 psi operating margin to the TSlimit.The Inservice Testing (IST) program establishes pump alert and action ranges as afunction of degradation from baseline.

The IST program sets alert and action limits forRHR pump degradation at 95% (alert range) and 93% (action range) of the reference LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS.,-

Enclosure 1 .Page 4 of 19pump curve. Figure 2 provides a graphical depiction of a typical pump curve, limits, andmargins.

In general, these ranges should provide indication of degradation prior topump performance falling below minimum analytical or TS limits. However, forRHR-P-2A and RHR-P-2B, this is not the case. The IST trending program ranges donot correlate or provide indication of degradation prior to exceeding TS limits.The close margir tpoft TS limits makes monitoring for actual degradation difficult, sinceinstrumentation c plibration uncertainties become a significant part of the data scatter.Trending becomE s very difficult with thle-rnrgin of instrument inaccuracy larger than thenormal operating margin. As a result, aCtual'pump degradation is masked. In addition, the small operati g margin, has made RHR ahdPRCS pump replacement andmaintenance diffi nult. ..* c.".Additionally; the equired.

performance Window for the RHpumps is narrow. It isbounded on the I w end by the TS minimum.

flow limit 745ýp gpm) and on thehigh. sidi:by the analtical flow limit (8100 gpm). Execuiting-thitý.

propos-ed change to g window larger will allow0for a morey ssah t,ys pmqtrý approachofture pump mainteman e, repair, and repla-cemetit

.or. futureThis LAR proposes a resolution.toitlbe issues o9ýsall operatingmarins thebaseline pump c e and theT.S lihmit and nrfowv required pe'rfbrnlancehindows.

Reassessing and reclaiming design ,ma.jin, as proposed bytIhis7,,LARi will address alegacy design issue that hasursitited in operftional and maintenance estrictions.

For RHRJLPCI

& LPCSthere are accidentRange of Normal Operations

.operating points for each,ýoe ratp ontsfr ahbut no 'range' of normaloperations Operating'Margin operating Limit (Ts Limi) ..." " -. '.., .'Design MarginAnalyzed Design Limit(LOCA Analysis Requirement)

Analytical MarginUltimate Capability.

FIGURE 1. MARGIN MODEL LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 5 of 19flow (PM) bFIGURE 2. TYPICAL PUMP CURVE, LIMITS AND MARGINS2.3 Description of Proposed Columbia Technical Specification ChangesEnergy Northwest has performed a detailed assessment of the issue and contracted with GE Hitachi Nuclear Energy (GEH) to provide an analysis to support lowering theLPCS and LPCI TS required flow rates. This analysis, coupled with previous

analyses, supports the following changes to the TS:SR 3.5.1.4 and SR 3.5.2.5:" LPCS Flow Rate is changed from 6,350 gpm to 6,200 gpm." LPCI Flow Rate is changed from 7,450 gpm to 7,200 gpm.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 6 of 19Additionally, the following editorial changes are proposed in SRs 3.5.1.4 and 3.5.2.5:* LPCI Differential Pressure unit is changed from psig to psid.* HPCS Differential Pressure unit is changed from psig to psid.The correct units of psid were approved at Amendment 149 but were inadvertently changed at Amendment 225, which consisted, in part, of the conversion of the entire TSfrom Word Perfect to Microsoft Word.3.0 TECHNICAL EVALUATION 3.1 LOCA ECCS AnalysisThe power uprate LAR (Reference

1) was submitted in 1993. The request wasapproved in 1995 under Amendment 137 (Reference 2). Part of this submittal includedNEDC-32115P, "SAFER/GESTR-LOCA" (Reference 3), which utilized theSAFER/GESTR-LOCA evaluation methodology to demonstrate conformance with theECCS acceptance criteria of 10 CFR 50.46. The approved application methodology consists of three essential parts. First, potentially limiting LOCA cases are determined by applying realistic (nominal) analytical models across the entire break spectrum.

Second, limiting LOCA cases are-analyzed with an Appendix K model (inputs andassumptions) that incorporates all the required features of 10 CFR 50 Appendix K. Forthe most limiting cases, a Licensing Basis Peak Cladding Temperature (PCT) iscalculated based on the nominal POT with an adder to account statistically for thedifferences between the nominal and Appendix K assumptions.

Finally, a statistically derived Upper Bound POT is calculated to demonstrate the conservatism of theLicensing Basis PCT. The resulting Licensing Basis PCT conforms to all therequirements of 10 CFR50.46 and Appendix K.In the license amendment request supporting the transition to Global Nuclear Fuel'sGE14 fuel design (References 4 and 5), Energy Northwest confirmed that theSAFER/GESTR-LOCA analysis continues to be the basis for the 10 CFR 50.46 LOCAanalysis.

Amendment 211 (Reference

6) was issued by the NRC in May 2009 andstates that the analysis methodology used by the licensee for the LOCA analysis is theNRC approved SAFER/GESTR-LOCA evaluation model.The above analyses utilized reduced analytical flow rates:* RHR/LPCI:

6,713 gpm with 26 psid between the reactor pressure vessel (RPV)and suppression pool" LPCS 5,625 gpm with 128 psid between the RPV and suppression pool LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 7 of 19As such, the ECCS-LOCA fuel analysis assessment bounds the proposed TS LPCI flowrate of 7,200 gpm and the proposed TS LPCS flow rate of 6,200 gpm.3.2 LOCA Containment AnalysisThe power uprate LAR also included NEDC-32141P, "Power Uprate with ExtendedLoad Line Limit Safety Analysis for WNP-2" (Reference 7), which summarized theevaluations performed to justify uprating the licensed thermal power to 3,486 MWt withan expanded operating domain. Section 4.1 of this report discussed the containment system performance.

As discussed in section 4.1.1.1 of NEDC-32141P, the analysiswas performed at 3,702 MWt using a more realistic decay heat table based on ANS 5.1-1979 decay heat and with a lower service water temperature of 90OF vs. 950F. As partof the detailed analysis conducted in support of the power uprate (Reference 8), GEHconducted sensitivity studies to quantify the effect of initial containment pressure on thecontainment response.

The containment response was analyzed at power uprateconditions with a 2 psig initial containment pressure as compared to a nominal value of0.7 psig assumed for the other cases. The peak drywell pressure and temperature increased by 2.6 psi and 30F, respectively.

The peak drywell-to-wetwell differential pressure was unaffected by the containment initial pressure increase.

Details of theexisting LOCA containment analysis are provided in FSAR Section 6.2 (Reference 9).Information on ECCS and containment cooling system parameters used in the existingcontainment analysis is contained in FSAR Table 6.2-2. As documented in this table,an analysis flow rate of 7,067 gpm was assumed for RHRFLPCI flow rate.Subsequently, in 2000, Energy Northwest was notified by GE Nuclear Energy of anincrease in the analyzed peak suppression pool temperature

(+0.50F) due to areassessment of the decay heat curve. The resultant peak suppression pooltemperature is 204.50F. Updated power uprate results are tabulated in FSAR Tables6.2-5 and 6.2-6.In order to support a change to the TS flow rates Energy Northwest contracted withGEH to perform a design basis accident (DBA) LOCA containment analysis and toperform assessments of all other RHR modes of operation affected by the proposedreduction in flow rates. The DBA-LOCA containment analysis was reevaluated tosupport lowering RHR/LPCI and LPCS Technical Specification flow rates and toevaluate GE Safety Communication (SC) 06-01, 'Worst Single Failure for Suppression Pool Temperature Analysis,"

January 19, 2006 (Reference 10).The revised DBA LOCA containment analysis (hereinafter referred to as "minimumECCS flow containment analysis")

and other RHR modes potentially affected by thereduced flow rates are addressed in GEH proprietary report NEDC-33813P, "Technical Specification Change Support for RHR/LPCI and LPCS Flow Rate Long-Term LOCAContainment Response and ECCS/Non-LOCA Evaluations" (Reference

11) and thenon-proprietary version of the report, NEDO-33813 (Reference 12). These reports are LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 8 of 19included with this LAR as Enclosures 2 and 3, respectively.

The analyses that do notcredit LPCI/LPCS/RHR flow (such as the containment short term analyses including dynamic loads and sub-compartment pressurization) are not affected by the reduction inflow rates.As documented in Enclosure 2, the GEH computer code SHEX was used to analyze thelong-term LOCA containment response for the minimum ECCS flow containment analysis.

The SHEX application methodology is documented in NEDO-10320, "The GEGeneral Electric Pressure Suppression Containment System Analytical Model"(Reference 13), and NEDO-20533,

'The General Electric Mark III Pressure Suppression Containment System Analytical Model" (Reference 14). This methodology was alsoutilized for the containment analysis performed for power uprate.Changes to key input parameters for the minimum ECCS flow containment analysisfrom those used in the power uprate analysis are listed in Table 1 and discussed below.TABLE 1 SUMMARY OF REVISED INPUT PARAMETERS Minimum ECCSParameter Units Power Uprate Flow Containment Analysis AnalysisContainment Cooling System" Before 600 sec -2 LPCI / 0 LPCS gpm 14,134/0 13,426/0" After 600 sec -1 LPCI / 0 LPCS gpm 7,067 / 0 6,713/0Reactor Power MWt 3,702 3,556ANS 5.1-1979

+ 2aDecay Heat ANS 5.1-1979 S 636with SIL 63685 for 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />sSW Temperature OF 90 te 90then 90RHR Heat Exchanger K value per Btu/sec-289 Reduced, variableloop OF from 284.5 to 288.8Time at which MSIVs are Fully Sec 3.5 3.0ClosedDrywell Relative Humidity

% 50 20Drywell Temperature OF 135 150The power uprate analysis assumed an initial power level of 3,702 MWt. Thispower corresponds to 102% of 3629 MWt. The analysis power was chosen tosupport a future uprate to 3629 MWt and bounds a power uprate to 3486 MWt(current licensed thermal power.) The minimum ECCS flow containment analysisassumed an initial power level of 3,556 MWt. This power corresponds to 102%of 3486 MWt. The decay heat contribution has been increased by 2a andactivation and actinide energies added per GE Service Information Letter (SIL)

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 9 of 19636, "Additional Terms Included in Reactor Decay Heat Calculations" (Reference 15)." The power uprate analysis assumed a constant value of 90'F for SWtemperature.

The SW temperature is limited by TS to 770F pre-accident.

Theultimate heat sink analysis (UHS) (Reference

16) shows that SW temperature does not exceed 85°F during the first ten hours following the LOCA. As such,SW temperature was assumed to be 850F for the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and 90°Fthereafter.

For additional discussion of the UHS analysis, see section 3.4 below.The RHR heat exchanger K-value is derived from an analysis of RHR heatexchanger capability based on reduced RHR flow rate, reduced SW flow rate,and combinations of SW and suppression pool temperatures (Reference 17)." The power uprate analysis assumed MSIV closure started at 0.5 seconds afterthe start of the accident.

The new analysis assumes MSIV closure time starts at0.0 seconds after the start of the accident, which increases containment heldenergy." The minimum ECCS flow containment analysis assumes drywell humidity isconservatively reduced and accounts for possible instrument inaccuracies.

Drywell temperature is conservatively increased and accounts for possibleinstrument uncertainties.

As documented in Enclosure 2 and summarized in Table 2 below, the results of theminimum ECCS flow containment analysis are bounded by the results of thecontainment analysis performed for power uprate.TABLE 2 SUMMARY OF ANALYSIS RESULTS FOR CASE CPower MinimumParameter Units Uprate ECCS Flow FSAR DesignAnalysis Containment Parameters AnalysisPeak Drywell Pressure psig 37.4 35.3 45Peak Drywell Temperature OF 283 281 340Peak Suppression Chamber psig 31.3 30.3 45PressurePeak Suppression Pool 203.8 204.5Temperature, long term -24The minimum ECCS flow containment analysis also includes an evaluation of GE SC06-01. The post LOCA scenario postulates that all ECCS equipment is operational except for one failed RHR heat exchanger.

This scenario requires that two RHR/LPCIpumps and the LPCS pump be secured to maintain suppression pool temperature LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 10 of 19within the analyzed limit of 204.50F. The specific timeframe for this action is contained in Enclosure

2. This timeframe provides sufficient time for the operator to respond.

Inthe event that one train of SW flow is lost or not available, procedural requirements direct the operators to secure the operating LPCS and LPCI pump(s) that are notrequired for adequate core cooling or containment integrity.

Actions to secure thepumps can be completed from the control room.3.3 Non-LOCA AnalysesNon-LOCA events were assessed to determine the effect of the TS RHR/LPCI andLPCS flow rate changes.

The results are provided in Enclosure 2 and show that thereduction in ECCS flow rates has no adverse effect on these events.3.4 UHS AnalysisFSAR Section 9.2.5 describes Columbia's UHS and the system and thermalperformance models. An analysis (Reference

18) was performed to determine theimpact of the reduction in LPCS and LPCI flow rates on peak SW spray pondtemperature.

LPCS Flow Rate Reduction:

LPCS pump heat load is a direct input to the suppression pool. Thus, it is conservative to continue to assume the full 6,350 gpm LPCS flow rate.Therefore, the change to LPCS flow has no effect on the results of the analysis.

LPCI Flow Rate Reduction:

The reduction in LPCI flow rate was analyzed to quantifythe effect on peak pond temperature.

It was determined that the change in RHR flowrate only affects the efficiency equation and results in a decrease in peak pondtemperature in the 4th decimal place. Thus, the change in RHR/LPCI flow rate does notresult in a change to the FSAR reported peak spray pond temperature.

Since the resultis a decrease in pond temperature, it is conservative to continue to assume the ratedflow of 7,450 gpm for LPCI/RHR in the UHS analysis.

SW Temperature Inputs to Minimum ECCS Flow Containment Analysis:

The revisedSW temperature values used in the minimum ECCS flow containment analysis moreaccurately reflect postulated accident conditions based on UHS analysis (Reference 16). The containment analysis assumes a SW temperature of 850F for the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.The UHS analysis assumes an initial SW temperature of 77°F, which is based on TS3.7.1, and predicts a SW spray pond temperature of 82.90F at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Thecontainment analysis then assumes a SW temperature of 90°F after 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. TheUHS analysis predicts a SW spray pond temperature of 89.5°F based on the worst caseanalysis.

Thus, the inputs to the minimum ECCS flow containment analysis bound thevalues predicted by the UHS analysis.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 11 of 193.5 Impacted Columbia Technical Specification SectionsThese analyses changes are reflected in the ECCS Technical Specification Surveillance Requirements as follows:* LPCI Flow Rate is changed from 7,450 gpm to 7,200 gpm" LPCS Flow Rate is changed from 6,350 gpm to 6,200 gpmThe LPCI TS change from 7,450 gpm to 7,200 gpm represents a 3% decrease whereasthe analytical decrease from 7,067 gpm to 6,713 gpm is a 5% decrease.

The LPCS TSchange from 6,350 gpm to 6,200 gpm represents a 2% decrease whereas the analytical decrease from 6,250 gpm to 5,625 gpm is a 10% decrease.

The difference between the analytical flow rate and the TS limiting flow rate represent margin to account for instrument uncertainty and potential variation in supply voltageand frequency.

The frequency variation

(+/- 2% for supply frequency),

voltage variation

(+/- 0.6% for supply voltage),

and instrument uncertainties

(+/- 2.5%) were combined insuch a manner as to produce the lowest, most conservative flow rates. Wheninstrument uncertainty and potential variation in supply voltage and frequency arefactored in, there is a difference of 151 gpm for LPCI and 194 gpm for LPCS betweentheir respective adjusted analysis flow rate and the TS limiting flow rate.3.6 Impact on Submittals under Review by NRCThe NRC is presently reviewing Energy Northwest's LAR to transition to the AveragePower Range Monitor (APRM) / Rod Block Monitor (RBM) Technical Specifications (ARTS) / Maximum Extended Load Line Limit Analysis (MELLLA) operation along withinstallation of the GEH Power Range Neutron Monitor (PRNM) system (Reference 19).The GEH evaluation scope contains an assessment of the impact of this change on theARTS/MELLLA analysis.

Conclusions are documented in Enclosure 2.4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements 4.1.1 10 CFR 50.46, 10 CFR 50 Appendix KThe acceptance criteria for ECCS performance include the following:

1. Peak cladding temperature.

The calculated maximum fuel element claddingtemperature shall not exceed 2,2000F.2. Maximum cladding oxidation.

The calculated total oxidation of the cladding shallnowhere exceed 0.17 times the total cladding thickness before oxidation.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 12 of 193. Maximum hydrogen generation.

The calculated total amount of hydrogengenerated from the chemical reaction of the cladding with water or steam shallnot exceed 0.01 times the hypothetical amount that would be generated if all themetal in the cladding cylinders surrounding the fuel, excluding the claddingsurrounding the plenum volume, were to react.4. Coolable geometry.

Calculated changes in core geometry shall be such that thecore remains amenable to cooling.5. Long-term cooling.

After any calculated successful initial operation of the ECCS,the calculated core temperature shall be maintained at an acceptably low valueand decay heat shall be removed for the extended period of time required by thelong-lived radioactivity remaining in the core.The above requirements are met and bounded by the analyses presented in FSARSection 6.3. The required minimum flow rates proposed in SR 3.5.1.4 and 3.5.2.5bound the analytical assumptions utilized in the ECCS LOCA fuel analyses.

Conservative analytical assumptions ensure that both short-term injection/cooling andlong-term cooling maintain previously approved safety margins.4.1.2 10 CFR 50 Appendix A General Design Criteria (GDC)The relevant GDCs are discussed below:Criterion 34-Residual heat removal A system to remove residual heat shall beprovided.

The system safety function shall be to transfer fission product decay heat andother residual heat from the reactor core at a rate such that specified acceptable fueldesign limits and the design conditions of the reactor coolant pressure boundary are notexceeded.

The RHR system provides the means to remove decay heat and residual heat from thenuclear system so that refueling and nuclear system servicing can be performed.

Themajor equipment of the RHR system consists of heat exchangers cooled by the SWsystem and main system pumps. The equipment is connected by associated valves andpiping. Additionally, there are controls and instrumentation provided for proper systemoperation.

The analysis provided in Enclosure 2 shows that the reduction in ECCS flowrates has no adverse effect on the ability of RHR to provide residual heat removal.Criterion 35-Emergency core cooling A system to provide abundant emergency corecooling shall be provided.

The system safety function shall be to transfer heat from thereactor core following any loss of reactor coolant at a rate such that (1) fuel and claddamage that could interfere with continued effective core cooling is prevented and (2)clad metal-water reaction is limited to negligible amounts.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 13 of 19The LPCS and LPCI systems are an integral part of the ECCS and provide redundancy and diversity in meeting the functional requirements of GDC 35. The systems areprovided to replace reactor vessel water inventory and to supply spray cooling of thecore following large pipe breaks in which the core may be uncovered.

The primarysafety function is therefore to deliver sufficient spray or flooding to each fuel bundle inthe core to prevent excessive fuel clad temperature following loss-of-coolant conditions.

The design is coordinated with the total ECCS in such a manner that for all rates ofcoolant loss from the primary reactor system the core is adequately cooled. Therequired minimum flow rates proposed in SR 3.5.1.4 and 3.5.2.5 bound the analytical assumptions utilized in the ECCS LOCA fuel analyses.

Criterion 37-Testing of Emergency Core Cooling System The emergency core coolingsystem shall be designed to permit appropriate periodic pressure and functional testingto assure (1) the structural and leak tight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of thesystem as a whole and, under conditions as close to design as practical, theperformance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer betweennormal and emergency power sources, and the operation of the associated coolingwater system.The LPCS and LPCI systems are an integral part of the ECCS, and are required tomeet the criteria specified in GDC 37. The systems are tested in accordance with theTS SRs in Specification 3.5.1 and 3.5.2. The required minimum flow rates specified inthe proposed SR 3.5.1.4 and 3.5.2.5 bound the analytical assumptions utilized in theECCS LOCA fuel and containment analyses.

Criterion 38-Containment heat removal A system to remove heat from the reactorcontainment shall be provided.

The system safety function shall be to reduce rapidly,consistent with the functioning of other associated

systems, the containment pressureand temperature following any loss-of-coolant accident and maintain them at acceptably low levels.The RHR system is designed specifically to perform this function.

The redundant coolant loops A and B are served by separate emergency power divisions, and eachloop contains a heat exchanger capable of removing the necessary heat to keepcontainment conditions (pressure and temperature) within design values. The analysisprovided in Enclosure 2 shows that the results of the minimum ECCS flow containment analysis are bounded by the power uprate analysis and do not exceed the designvalues specified in the FSAR.Criterion 40-Testing of Containment Heat Removal System The containment heatremoval system shall be designed to permit appropriate periodic pressure and LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 14 of 19functional testing to assure (1) the structural and leak tight integrity of its components, (2) the operability and performance of the active components of the systems, and (3)the operability of the system as a whole, and under conditions as close to the design aspractical, the performance of the full operational sequence that brings the system intooperation, including operation of applicable portions of the protection system, thetransfer between normal and emergency power sources, and the operation of theassociated cooling water system.The RHR containment spray and cooling function is required to meet the criteriaspecified in GDC 40. The system is tested in accordance with the applicable TS SRs inSpecifications 3.6.1.5 and 3.6.2.3.

The flow rate specified in SR 3.6.2.3.2 of > 7,100gpm bounds the analytical assumptions utilized in the minimum ECCS flow containment analysis.

Criterion 44-Cooling water A system to transfer heat from structures,

systems, andcomponents important to safety, to an ultimate heat sink shall be provided.

The systemsafety function shall be to transfer the combined heat load of these structures, systems,and components under normal operating and accident conditions.

The safety-related cooling water system is the SW system, which supplies cooling forthe RHR, LPCS, High Pressure Core Spray (HPCS) system, FPC system, emergency diesel generators, and the essential

heating, ventilation and air conditioning (HVAC)systems.

The redundant SW systems are open loop systems which transfer heat fromstructures,

systems, and safety-related components to the UHS. The UHS, whichconsists of two man-made Seismic Category I spray ponds, is designed to withstand extreme natural phenomena.

The impact of the reduced LPCS and LPCI flow rates onthe UHS analysis was evaluated to determine the impact on peak SW spray pondtemperature.

The reduction in flow rates does not increase the peak SW spray pondtemperature.

The inputs to the minimum ECCS flow containment analysis bound thevalues predicted by the UHS analysisCriterion 50-Containment desiQn basis The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall bedesigned so that the containment structure and its internal compartments canaccommodate, without exceeding the design leakage rate and with sufficient margin,the calculated pressure and temperature conditions resulting from any loss-of-coolant accident.

The analysis provided in Enclosure 2 shows that the results of the minimum ECCS flowcontainment analysis are bounded by the power uprate analysis and do not exceed thedesign values specified in the FSAR.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 15 of 194.2 Applicable Regulatory GuidanceNUREG-0800, Standard Review Plan (SRP) Section 6.2.1.1.C, "Pressure-Suppression Type BWR Containments,"

states that the peak calculated values of pressure andtemperature for the drywell and wetwell should not exceed the respective design values.The analysis provided in Enclosure 2 shows that the results of the minimum ECCS flowcontainment analysis do not exceed the design values specified in the FSAR.5.0 PRECEDENT The GEH evaluation methodology SAFER/GESTR-LOCA is used to analyze ECCSperformance.

NRC approval of the SAFE R/GESTR-LOCA evaluation methodology isdocumented in Reference

20. Approval of this methodology for use at Columbia isdocumented in References 2 and 6. The GEH computer code SHEX is used to analyzethe long-term DBA LOCA containment response.

References 13 and 14 document theSHEX application methodology.

Reference 21 documents the NRC acceptance of theapplication of SHEX for containment analyses.

Approval of this methodology for use atColumbia is documented in Reference 2.6.0 SIGNIFICANT HAZARDS CONSIDERATION Energy Northwest has evaluated whether or not a significant hazards consideration isinvolved with the proposed amendment by focusing on the three standards set forth in10 CFR 50.92, "Issuance of amendment,"

as discussed below:1) Does the proposed amendment involve a significant increase in the probability orconsequences of an accident previously evaluated?

Response:

No.The proposed change would lower the required LPCI and LPCS flow rates in SR3.5.1.4 and 3.5.2.5.

The requested changes do not serve as initiators of anyColumbia accident previously evaluated.

The existing ECCS-LOCA fuel analysis ofrecord utilizes reduced analytical flow rates that bound the proposed TS LPCI andLPCS flow rates. The analysis demonstrates compliance with the ECCSacceptance criteria in 10 CFR 50.46. The new minimum ECCS flow containment analysis also utilizes reduced analytical flow rates that bound the proposed TSLPCI and LPCS flow rates. This analysis demonstrates that the results of theanalysis do not exceed the design values specified in the FSAR, which isconsistent with the acceptance criteria specified in SRP 6.2.1.1.C.

The accidentprobabilities are unaffected and the consequences remain unchanged.

Therefore there is no significant increase in the probability or consequences of anaccident previously evaluated.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 16 of 192) Does the proposed amendment create the possibility of a new or different kind ofaccident from any accident previously analyzed?

Response:

No.There are no postulated

hazards, new or different, contained in this amendment.

Analysis has determined that these changes have been bounded by previousevaluations.

Therefore, the proposed change does not create the possibility of a new ordifferent kind of accident from any accident previously evaluated.

3) Does the proposed amendment involve a significant reduction in a margin ofsafety?Response:

No.The proposed changes lower the TS SR flows for LPCI and LPCS by 3% and 2%,respectively.

The analytical values for the LPCI and LPCS flows were reduced by5% and 10%, respectively, to ensure no margin of safety was impacted.

To ensurea bounding calculation, the minimum ECCS flow containment analysis wasperformed with conservative assumptions and using NRC approved methodologies previously accepted for use at Columbia by the NRC. The proposed TS limitingflow rates provide adequate margin to the analytical limits accounting for worst-case instrument uncertainty and potential variation in supply voltage andfrequency.

Therefore, the proposed change does not involve a significant reduction in themargin of safety.Based on the above, Energy Northwest concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forthin 10 CFR 50.92(c),

and, accordingly, a finding of "no significant hazardsconsideration" is justified.

7.0 CONCLUSION

S Based on the considerations discussed above: (1) there is reasonable assurance thatthe health and safety of the public will not be endangered by operation in the proposedmanner, (2) such activities will be conducted in compliance with the applicable regulations as identified herein, and (3) the issuance of the amendment will not beinimical to the common defense and security or to the health and safety of the public.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 17 of 198.0 ENVIRONMENTAL CONSIDERATION Energy Northwest has determined that the proposed amendment would changerequirements with respect to installation or use of a facility component located withinColumbia's restricted area, as defined in 10 CFR 20, or would change an inspection orsurveillance requirement.

Energy Northwest has evaluated the proposed change andhas determined that the change does not involve, (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of anyeffluents that may be released

offsite, or (iii) a significant increase in individual orcumulative occupational radiation exposure.

Accordingly, the proposed change meetsthe eligibility criteria for categorical exclusion in accordance with 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement orenvironmental assessment need be prepared in connection with the proposedamendment.

9.0 REFERENCES

1. Letter G02-93-180, JV Parrish (Washington Public Power Supply System) toNRC, WNP-2 Operating License NPF-21 Request for Amendment to the FacilityOperating License and Technical Specifications to Increase Licensed PowerLevel From 3323 MWt to 3486 MWt With Extended Load Line Limit and Changein Safety Relief Valve Setpoint Tolerance, dated July 9, 1993.2. Letter, JW Clifford (NRC) to JV Parrish (Washington Public Power SupplySystem),

Issuance of Amendment for the Washington Public Power SupplySystem Nuclear Project No. 2 (TAC NOS. M87076 and M88625),

dated May 2,1995. (ADAMS Accession No. ML022120154).

3. GE Nuclear Energy, NEDC-32115P, Washington Public Power Supply System,Nuclear Project 2, SAFER/GESTR-LOCA, Loss-of-Coolant Accident

Analysis, Revision 2, July 1993.4. Letter G02-08-108, SK Gambhir (Energy Northwest) to NRC, LicenseAmendment Request for Changes to Technical Specifications Involving CoreOperating Limits Report and Scram Time Testing, dated July 16, 2008.5. Letter G02-09-050, SK Gambhir (Energy Northwest) to NRC, Supplemental Response to Request for Additional Information (RAI) Regarding LicenseAmendment Request Involving Core Operating Limits Report and Scram TimeTesting, dated March 19, 2009.6. Letter, CF Lyon (NRC) to JV Parrish (Energy Northwest),

Columbia Generating Station -Issuance of Amendment Re: Core Operating Limits Report and ScramTime Testing (TAC No. MD9247),

dated May 5, 2009.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 18 of 197. GE Nuclear Energy, NEDC-32141 P, Power Uprate with Extended Load LineLimit Safety Analysis for WNP-2, June 1993.8. GE Nuclear Energy, GE-NE-208-17-0993, WNP-2 Power Uprate Project NSSSEngineering Report, Revision 1, December 1994.9. Energy Northwest, Columbia Generating

Station, Final Safety Analysis ReportAmendment 61.10. General Electric (GE) Safety Communication (SC) 06-01 Worst Single Failure forSuppression Pool Temperature

Analysis, January 19, 2006.11. GE Hitachi Nuclear Energy, NEDC-33813P, Technical Specification ChangeSupport for RHR/LPCI and LPCS Flow Rate Long-Term LOCA Containment Response and ECCS/Non-LOCA Evaluations, Revision 2, September 2013.12. GE Hitachi Nuclear Energy, NEDO-33813, Technical Specification ChangeSupport for RHR/LPCI and LPCS Flow Rate Long-Term LOCA Containment Response and ECCS/Non-LOCA Evaluations, Revision 2, September 2013.13. NEDO-10320, The GE General Electric Pressure Suppression Containment System Analytical Model, March 1971.14. NEDO-20533, The General Electric Mark Ill Pressure Suppression Containment System Analytical Model, June 1974.15. GE Nuclear Energy Service Information Letter (SIL) Number 636, Additional Terms Included in Reactor Decay Heat Calculations, Revision 1, June 6, 2001.16. Energy Northwest Calculation, ME-02-92-41, Ultimate Heat Sink Analysis, Revision 6.17. Energy Northwest Calculation, ME-02-93-20, Calculation for RHR Operation atReduced Flowrates, CMR 11549.18. Energy Northwest Calculation, ME-02-92-41, Ultimate Heat Sink Analysis, Calculation Modification Record (CMR) 11561.19. Letter G02-12-017, BJ Sawatzke (Energy Northwest) to NRC, LicenseAmendment Request to Change Technical Specifications in support of PRNM /ARTS/MELLLA Implementation, dated January 31, 2012.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 19 of 1920. Letter, CO Thomas (NRC) to JF Quirk (GE), Acceptance for Referencing ofLicensing Topical Report NEDE-23785, Revision 1, Volume Ill (P), 'The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant

Accident, June 1, 1984.21. Letter, Ashok Thadani (NRC) to GL Sozzi (GE), Use of SHEX Computer Programand ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature

Analysis, July 13, 1993.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure IPage 1 of 1Attachment IProposed Columbia Technical Specification Changes (Mark-Up)

ECCS -Operating 3.5.1SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.2-.... .... -.. .. .. O T E ...- .....Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactorsteam dome pressure less than 48 psig in MODE 3,if capable of being manually realigned and nototherwise inoperable.

Verify each ECCS injectiornspray subsystem manual, power operated, and automatic valve In theflow path, that is not locked, sealed, or otherwise secured in position, Is in the correct position.

31 daysSR 3.5.1.3 Verify ADS accumulator backup compressed gas 31 dayssystem average pressure In the required bottles is22200 psig.SR 3.5.1.4 Verify each ECCS pump develops the specified flow In accordance rate with the specified differential pressure between with the Inservice reactor and suction source. Testing ProgramDIFFERENTIAL PRESSUREBETWEENREACTOR ANDSYSTEM FLOW RATE SUCTION SOURCELPCS > gpm > 128 psidLPCI > 746F-7200 gpm > 26 psidgHPCS > 6350 gpm ! 200 psidgSR 3.5.1.5 NOTE ..Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem 24 monthsactuates on an actual or simulated automatic initiation signal.Columbia Generating Station3.5.1-4Amendment No. 460,246 225 ECCS -Shutdown3.5.2SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify, for each required ECCS injection/spray 31 dayssubsystem, the piping is filled with water from thepump discharge valve to the Injection valve.SR 3.5.2.4 ---NOTEOne low pressure coolant injection (LPCI)subsystem may be considered OPERABLE duringalignment and operation for decay heat removal, ifcapable of being manually realigned and nototherwise Inoperable.

Verify each required ECCS Injection/spray 31 dayssubsystem manual, power operated, and automatic valve in the flow path, that Is not locked, sealed, orotherwise secured in position, is in the correctposition.

SR 3.5.2.5 Verify each required ECCS pump develops the In accordance specified flow rate with the specifeld differential with the Inservice pressure between reactor and suction source. Testing ProgramDIFFERENTIAL PRESSUREBETWEENREACTOR AND-F_.MEO.,RATE SUCTION SOURCELPCS > 36.0-6200 gpm 128 psidLPCI > 7 4rag7200 gpm > 26 psid§HPCS > 6350 gpm > 200 psidgSR 3.5.2.6 -----NOTE-Vessel injection/spray may be excluded.

Verify each required ECCS injection/spray 24 monthssubsystem actuates on an actual or simulated automatic initiation signal.Columbia Generating Station3.5.2-3Amendment No. 414,206 225 LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORECOOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1Page 1 of IAttachment 2Proposed Columbia Technical Specification Changes (Re-Typed)

ECCS -Operating 3.5.1SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.2.--NOTE --.Low pressure coolant Injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactorsteam dome pressure less than 48 psig in MODE 3,if capable of being manually realigned and nototherwise inoperable.

Verify each ECCS injection/spray subsystem manual, power operated, and automatic valve in theflow path, that is not locked, sealed, or otherwise secured in position, Is in the correct position.

31 daysSR 3.5.1.3 Verify ADS accumulator backup compressed gas 31 dayssystem average pressure in the required bottles is 2200 psig.SR 3.5.1.4 Verify each ECCS pump develops the specified flow In accordance rate with the specified differential pressure between with the Inservice reactor and suction source. Testing ProgramDIFFERENTIAL PRESSUREBETWEENREACTOR ANDSYSTEM FLOW RATE SUCTION SOURCELPCS 6200 gpm > 128 psidLPCI 7200 gpm > 26 psidHPCS _> 6350 gpm 200 psidSR 3.5.1.5Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem 24 monthsactuates on an actual or simulated automatic initiation signal.Columbia Generating Station3.5.1-4Amendment No. 460,2-0 225 ECCS -Shutdown3.5.2SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify, for each required ECCS Injection/spray 31 dayssubsystem, the piping is filled with water from thepump discharge valve to the Injection valve.SR 3.5.2.4 .......-NOTEOne low pressure coolant injection (LPCI)subsystem may be considered OPERABLE duringalignment and operation for decay heat removal, ifcapable of being manually realigned and nototherwise inoperable.

Verify each required ECCS injection/spray 31 dayssubsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, orotherwise secured in position, is in the correctposition.

SR 3.5.2.5 Verify each required ECCS pump develops the In accordance specified flow rate with the specified differential with the Inservice pressure between reactor and suction source. Testing ProgramDIFFERENTIAL PRESSUREBETWEENREACTOR ANDSYSIEM FLOWM RIE SUCTION SOURCELPCS >6200 gpm _ 128 psidLPCI >7200 gpm _ 26 psidHPCS >6350 gpm >200 psidSR 3.5.2.6 NOTE--Vessel injection/spray may be excluded.

Verify each required ECCS Injection/spray 24 monthssubsystem actuates on an actual or simulated automatic Initiation signal.Columbia Generating Station3.5.2-3Amendment No. 469,205 225