ML13316A009

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 Station W /u P.O. Box 968, PE04 L~J Richiarid, WA 993-20968 Ph.

F. 509.377.4150 Proprietary

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

October 31, 2013 G02-13-151 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Subject COLUMBIA GENERATING STATION, DOCKET NO. 50"397 LUCENSE 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 to revise 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 Low Pressure Coolant Injection (LPCI) pump flows. This amendment is requested to increase 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 LOCA Containment Response and ECCS/Non-LOCA Evaluations".

GE Hitachi Nuclear Energy (GEH) considers certain information contained in Enclosure 2 to be proprietary and, therefore, requests that it be withheld from public disclosure in accordance with 10 CFR 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 the document, 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 the submittal.

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 CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Page 2 of 2 In accordance with 10 CFR 50.91, Energy Northwest is notifying the State of Washington of this amendment request by transmitting a copy of this letter and enclosures 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 on the date of this letter.Respectfully, A. L. Javonk Vice President, Engineering

Enclosures:

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

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 1 of 19 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION This evaluation supports a License Amendment Request (LAR) to lower the flow rates specified in Columbia Generating Station (Columbia)

Technical Specification (TS) 3.5.1 and 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 6,350 gpm to 6,200 gpm for LPCS and from 7,450 gpm to 7,200 gpm for RHR/LPCI at the same specified differential pressure between the reactor pressure vessel (RPV) and suppression pool (128 psid for LPCS and 26 psid for RHR/LPCI).

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

2.1 Background

2.1.1 LPCS The 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 the core 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 assisted by the Automatic Depressurization System (ADS), LPCS also provides protection for small breaks.The LPCS system consists of a single motor-driven centrifugal pump, a spray sparger in the reactor vessel above the core, piping and valves to convey water from the suppression pool to the sparger, and associated controls and instrumentation.

Low Pressure Core Spray is associated with Division 1.2.1.2 LPCI The LPCI mode is an operating mode of the Residual Heat Removal (RHR) system and is one of the systems in the ECCS network dedicated to assure that postulated LOCA consequences can be mitigated.

The LPCI mode delivers water to the core at low reactor pressures.

The primary purpose of LPCI is to provide inventory makeup following large pipe breaks. When assisted by ADS, LPCI also provides protection for small breaks.

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

For the LPCI mode of RHR, the three pumps deliver water from the suppression pool to the bypass region inside the shroud through three separate reactor vessel penetrations and cool the core by 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 and revert system valves to the LPCI lineup. The RHR system continues in the LPCI mode until the operator determines that another mode of operation is needed (such as containment cooling) and takes action to manually initiate that mode. LPCI will not be diverted to any other mode of operation until adequate core cooling is ensured. No operator 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 with Division 2.2.1.3 RHR Containment Heat Removal The RHR containment heat removal function is accomplished by the use of an operational mode of the RHR system. The purpose of this system is to prevent excessive 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 heat removal capability.

The RHR system's suppression pool cooling (SPC) and containment spray cooling (CSC) modes provide heat removal from the suppression pool and containment by pumping suppression pool water through the system's heat exchangers 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 the suppression pool. The drywell spray function also removes radioactive fission products from the containment atmosphere during a LOCA. Water from the Standby Service Water (SW) system is pumped through the heat exchanger tube side to remove heat from the process water.There are no signals which automatically initiate containment cooling; however, the SW system is automatically initiated by the same signals which start up the ECCS. To start RHR containment cooling after a LOCA resulting from a large break, the operator verifies that the normally open RHR heat exchanger isolation valves are open and then shuts the heat exchanger bypass valve. The rated containment cooling flow, 7,450 gpm, can be achieved through the LPCI line, the drywell spray line, or through the test line and wetwell spray line, which directs the heat exchanger discharge directly into the suppression pool. Thus, the design allows containment cooling simultaneously with core flooding or containment spray. If the break size is small enough to limit reactor LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 3 of 19 depressurization, the rated containment cooling flow cannot be established through the LPCI line. The operator must then direct the RHR containment cooling flow through the drywell spray line or through the test line; however, the operator will not divert LPCI flow away from the reactor until adequate core cooling is ensured. When directed by procedures, the operator may start drywell spray by shutting the LPCI injection valve and then opening the drywell spray valves. Similarly, the operator may divert the flow directly to the suppression pool by shutting the LPCI injection valve and then opening the test line valve.2.1.4 RHR Shutdown Cooling The RHR system's normal shutdown cooling mode removes reactor core decay and sensible heat from the primary reactor system to permit refueling and servicing.

This heat removal function is initiated manually after the reactor pressure has been reduced to less than 48 psig (295 0 F) by discharge of steam to the main condenser.

The RHR system's alternate shutdown cooling mode is utilized during normal plant operation and design basis events when the normal shutdown cooling mode is not available to remove reactor core decay and sensible heat. This heat removal function is safety related, initiated manually and pumps suppression pool water into the core and allows 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, the RHR system is also designed to be capable of being aligned to assist the Fuel Pool Cooling and Cleanup (FPC) system in maintaining the fuel pool temperature within acceptable limits. In this mode the system is designed to cool water drawn from the fuel pool by passing it through an RHR system heat exchanger and then discharge the water back to the fuel pool.2.2 Circumstances Necessitating the Change This LAR requests a TS change to redefine the operating margin for safety related LPCS and RHR/LPCI pumps.Historically, the plant has had little operating margin with these pumps. See Figure 1 for a representation of margins. The RHR pumps were tested prior to initial plant startup using the actual injection flow path. RHR-P-2B produced 7,500 gpm at 28 psid on 9/24/1983.

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

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

Enclosure 1 .Page 4 of 19 pump curve. Figure 2 provides a graphical depiction of a typical pump curve, limits, and margins. In general, these ranges should provide indication of degradation prior to pump performance falling below minimum analytical or TS limits. However, for RHR-P-2A and RHR-P-2B, this is not the case. The IST trending program ranges do not correlate or provide indication of degradation prior to exceeding TS limits.The close margir tpoft TS limits makes monitoring for actual degradation difficult, since instrumentation c plibration uncertainties become a significant part of the data scatter.Trending becomE s very difficult with thle-rnrgin of instrument inaccuracy larger than the normal operating margin. As a result, aCtual'pump degradation is masked. In addition, the small operati g margin, has made RHR ahdPRCS pump replacement and maintenance diffi nult. ..* c.".Additionally; the equired. performance Window for the RHpumps is narrow. It is bounded on the I w end by the TS minimum. flow limit 745ýp gpm) and on the high. 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. future This LAR proposes a resolution.toitlbe issues o9ýsall operatingmarins the baseline pump c e and theT.S lihmit and nrfowv required pe'rfbrnlancehindows.

Reassessing and reclaiming design ,ma.jin, as proposed bytIhis 7 ,,LARi will address a legacy design issue that hasursitited in operftional and maintenance estrictions.

For RHRJLPCI & LPCS there are accident Range of Normal Operations

.operating points for each,ýoe ratp ontsfr ah but no 'range' of normal operations Operating'Margin operating Limit (Ts Limi) ..." " -. '.., .'Design Margin Analyzed Design Limit (LOCA Analysis Requirement)

Analytical Margin Ultimate Capability.

FIGURE 1. MARGIN MODEL LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 5 of 19 flow (PM) b FIGURE 2. TYPICAL PUMP CURVE, LIMITS AND MARGINS 2.3 Description of Proposed Columbia Technical Specification Changes Energy Northwest has performed a detailed assessment of the issue and contracted with GE Hitachi Nuclear Energy (GEH) to provide an analysis to support lowering the LPCS 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 CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 6 of 19 Additionally, 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 TS from Word Perfect to Microsoft Word.3.0 TECHNICAL EVALUATION 3.1 LOCA ECCS Analysis The power uprate LAR (Reference

1) was submitted in 1993. The request was approved in 1995 under Amendment 137 (Reference 2). Part of this submittal included NEDC-32115P, "SAFER/GESTR-LOCA" (Reference 3), which utilized the SAFER/GESTR-LOCA evaluation methodology to demonstrate conformance with the ECCS 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 and assumptions) that incorporates all the required features of 10 CFR 50 Appendix K. For the most limiting cases, a Licensing Basis Peak Cladding Temperature (PCT) is calculated based on the nominal POT with an adder to account statistically for the differences between the nominal and Appendix K assumptions.

Finally, a statistically derived Upper Bound POT is calculated to demonstrate the conservatism of the Licensing Basis PCT. The resulting Licensing Basis PCT conforms to all the requirements of 10 CFR50.46 and Appendix K.In the license amendment request supporting the transition to Global Nuclear Fuel's GE14 fuel design (References 4 and 5), Energy Northwest confirmed that the SAFER/GESTR-LOCA analysis continues to be the basis for the 10 CFR 50.46 LOCA analysis.

Amendment 211 (Reference

6) was issued by the NRC in May 2009 and states that the analysis methodology used by the licensee for the LOCA analysis is the NRC 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 CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 7 of 19 As such, the ECCS-LOCA fuel analysis assessment bounds the proposed TS LPCI flow rate of 7,200 gpm and the proposed TS LPCS flow rate of 6,200 gpm.3.2 LOCA Containment Analysis The power uprate LAR also included NEDC-32141P, "Power Uprate with Extended Load Line Limit Safety Analysis for WNP-2" (Reference 7), which summarized the evaluations performed to justify uprating the licensed thermal power to 3,486 MWt with an 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 analysis was 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. 95 0 F. As part of the detailed analysis conducted in support of the power uprate (Reference 8), GEH conducted sensitivity studies to quantify the effect of initial containment pressure on the containment response.

The containment response was analyzed at power uprate conditions with a 2 psig initial containment pressure as compared to a nominal value of 0.7 psig assumed for the other cases. The peak drywell pressure and temperature increased by 2.6 psi and 3 0 F, respectively.

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

Details of the existing LOCA containment analysis are provided in FSAR Section 6.2 (Reference 9).Information on ECCS and containment cooling system parameters used in the existing containment 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 an increase in the analyzed peak suppression pool temperature

(+0.5 0 F) due to a reassessment of the decay heat curve. The resultant peak suppression pool temperature is 204.5 0 F. Updated power uprate results are tabulated in FSAR Tables 6.2-5 and 6.2-6.In order to support a change to the TS flow rates Energy Northwest contracted with GEH to perform a design basis accident (DBA) LOCA containment analysis and to perform assessments of all other RHR modes of operation affected by the proposed reduction in flow rates. The DBA-LOCA containment analysis was reevaluated to support lowering RHR/LPCI and LPCS Technical Specification flow rates and to evaluate 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 "minimum ECCS flow containment analysis")

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

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

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

The SHEX application methodology is documented in NEDO-10320, "The GE General 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 also utilized for the containment analysis performed for power uprate.Changes to key input parameters for the minimum ECCS flow containment analysis from those used in the power uprate analysis are listed in Table 1 and discussed below.TABLE 1

SUMMARY

OF REVISED INPUT PARAMETERS Minimum ECCS Parameter Units Power Uprate Flow Containment Analysis Analysis Containment 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 6,713/0 Reactor Power MWt 3,702 3,556 ANS 5.1-1979 + 2a Decay Heat ANS 5.1-1979 S 636 with SIL 636 85 for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> SW Temperature OF 90 te 90 then 90 RHR Heat Exchanger K value per Btu/sec- 289 Reduced, variable loop OF from 284.5 to 288.8 Time at which MSIVs are Fully Sec 3.5 3.0 Closed Drywell Relative Humidity % 50 20 Drywell Temperature OF 135 150 The power uprate analysis assumed an initial power level of 3,702 MWt. This power corresponds to 102% of 3629 MWt. The analysis power was chosen to support a future uprate to 3629 MWt and bounds a power uprate to 3486 MWt (current licensed thermal power.) The minimum ECCS flow containment analysis assumed 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 and activation and actinide energies added per GE Service Information Letter (SIL)

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 9 of 19 636, "Additional Terms Included in Reactor Decay Heat Calculations" (Reference 15)." The power uprate analysis assumed a constant value of 90'F for SW temperature.

The SW temperature is limited by TS to 77 0 F pre-accident.

The ultimate 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 85 0 F 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°F thereafter.

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 heat exchanger 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 after the start of the accident.

The new analysis assumes MSIV closure time starts at 0.0 seconds after the start of the accident, which increases containment held energy." The minimum ECCS flow containment analysis assumes drywell humidity is conservatively reduced and accounts for possible instrument inaccuracies.

Drywell temperature is conservatively increased and accounts for possible instrument uncertainties.

As documented in Enclosure 2 and summarized in Table 2 below, the results of the minimum ECCS flow containment analysis are bounded by the results of the containment analysis performed for power uprate.TABLE 2

SUMMARY

OF ANALYSIS RESULTS FOR CASE C Power Minimum Parameter Units Uprate ECCS Flow FSAR Design Analysis Containment Parameters Analysis Peak Drywell Pressure psig 37.4 35.3 45 Peak Drywell Temperature OF 283 281 340 Peak Suppression Chamber psig 31.3 30.3 45 Pressure Peak Suppression Pool 203.8 204.5 Temperature, long term -24 The minimum ECCS flow containment analysis also includes an evaluation of GE SC 06-01. The post LOCA scenario postulates that all ECCS equipment is operational except for one failed RHR heat exchanger.

This scenario requires that two RHR/LPCI pumps and the LPCS pump be secured to maintain suppression pool temperature LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 10 of 19 within the analyzed limit of 204.5 0 F. The specific timeframe for this action is contained in Enclosure

2. This timeframe provides sufficient time for the operator to respond. In the 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 not required for adequate core cooling or containment integrity.

Actions to secure the pumps can be completed from the control room.3.3 Non-LOCA Analyses Non-LOCA events were assessed to determine the effect of the TS RHR/LPCI and LPCS flow rate changes. The results are provided in Enclosure 2 and show that the reduction in ECCS flow rates has no adverse effect on these events.3.4 UHS Analysis FSAR Section 9.2.5 describes Columbia's UHS and the system and thermal performance models. An analysis (Reference

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

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 quantify the effect on peak pond temperature.

It was determined that the change in RHR flow rate only affects the efficiency equation and results in a decrease in peak pond temperature in the 4 th decimal place. Thus, the change in RHR/LPCI flow rate does not result in a change to the FSAR reported peak spray pond temperature.

Since the result is a decrease in pond temperature, it is conservative to continue to assume the rated flow of 7,450 gpm for LPCI/RHR in the UHS analysis.SW Temperature Inputs to Minimum ECCS Flow Containment Analysis:

The revised SW temperature values used in the minimum ECCS flow containment analysis more accurately reflect postulated accident conditions based on UHS analysis (Reference 16). The containment analysis assumes a SW temperature of 85 0 F 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 TS 3.7.1, and predicts a SW spray pond temperature of 82.9 0 F at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The containment 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 />. The UHS analysis predicts a SW spray pond temperature of 89.5°F based on the worst case analysis.

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

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 11 of 19 3.5 Impacted Columbia Technical Specification Sections These 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 gpm The LPCI TS change from 7,450 gpm to 7,200 gpm represents a 3% decrease whereas the analytical decrease from 7,067 gpm to 6,713 gpm is a 5% decrease.

The LPCS TS change 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 voltage and frequency.

The frequency variation

(+/- 2% for supply frequency), voltage variation (+/- 0.6% for supply voltage), and instrument uncertainties

(+/- 2.5%) were combined in such a manner as to produce the lowest, most conservative flow rates. When instrument uncertainty and potential variation in supply voltage and frequency are factored in, there is a difference of 151 gpm for LPCI and 194 gpm for LPCS between their respective adjusted analysis flow rate and the TS limiting flow rate.3.6 Impact on Submittals under Review by NRC The NRC is presently reviewing Energy Northwest's LAR to transition to the Average Power Range Monitor (APRM) / Rod Block Monitor (RBM) Technical Specifications (ARTS) / Maximum Extended Load Line Limit Analysis (MELLLA) operation along with installation 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 the ARTS/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 K The acceptance criteria for ECCS performance include the following:

1. Peak cladding temperature.

The calculated maximum fuel element cladding temperature shall not exceed 2,200 0 F.2. Maximum cladding oxidation.

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

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 12 of 19 3. Maximum hydrogen generation.

The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.4. Coolable geometry.

Calculated changes in core geometry shall be such that the core 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 value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.The above requirements are met and bounded by the analyses presented in FSAR Section 6.3. The required 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.Conservative analytical assumptions ensure that both short-term injection/cooling and long-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 be provided.

The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.The RHR system provides the means to remove decay heat and residual heat from the nuclear system so that refueling and nuclear system servicing can be performed.

The major equipment of the RHR system consists of heat exchangers cooled by the SW system and main system pumps. The equipment is connected by associated valves and piping. Additionally, there are controls and instrumentation provided for proper system operation.

The analysis provided in Enclosure 2 shows that the reduction in ECCS flow rates 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 core cooling shall be provided.

The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage 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 CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 13 of 19 The 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 are provided to replace reactor vessel water inventory and to supply spray cooling of the core following large pipe breaks in which the core may be uncovered.

The primary safety function is therefore to deliver sufficient spray or flooding to each fuel bundle in the 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 of coolant loss from the primary reactor system the core is adequately cooled. The required 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 cooling system shall be designed to permit appropriate periodic pressure and functional testing to 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 the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.The LPCS and LPCI systems are an integral part of the ECCS, and are required to meet the criteria specified in GDC 37. The systems are tested in accordance with the TS SRs in Specification 3.5.1 and 3.5.2. The required minimum flow rates specified in the proposed SR 3.5.1.4 and 3.5.2.5 bound the analytical assumptions utilized in the ECCS LOCA fuel and containment analyses.Criterion 38-Containment heat removal A system to remove heat from the reactor containment shall be provided.

The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and 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 each loop contains a heat exchanger capable of removing the necessary heat to keep containment conditions (pressure and temperature) within design values. The analysis provided 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 design values specified in the FSAR.Criterion 40-Testing of Containment Heat Removal System The containment heat removal system shall be designed to permit appropriate periodic pressure and LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 14 of 19 functional 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 as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.The RHR containment spray and cooling function is required to meet the criteria specified in GDC 40. The system is tested in accordance with the applicable TS SRs in Specifications 3.6.1.5 and 3.6.2.3. The flow rate specified in SR 3.6.2.3.2 of > 7,100 gpm bounds the analytical assumptions utilized in the minimum ECCS flow containment analysis.Criterion 44-Cooling water A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided.

The system safety 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 for the 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 from structures, systems, and safety-related components to the UHS. The UHS, which consists 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 on the UHS analysis was evaluated to determine the impact on peak SW spray pond temperature.

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

The inputs to the minimum ECCS flow containment analysis bound the values predicted by the UHS analysis Criterion 50-Containment desiQn basis The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, 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 flow containment analysis are bounded by the power uprate analysis and do not exceed the design values specified in the FSAR.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 15 of 19 4.2 Applicable Regulatory Guidance NUREG-0800, Standard Review Plan (SRP) Section 6.2.1.1.C, "Pressure-Suppression Type BWR Containments," states that the peak calculated values of pressure and temperature 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 flow containment 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 ECCS performance.

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

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

References 13 and 14 document the SHEX application methodology.

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

Approval of this methodology for use at Columbia is documented in Reference

2.6.0 SIGNIFICANT

HAZARDS CONSIDERATION Energy Northwest has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: 1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No.The proposed change would lower the required LPCI and LPCS flow rates in SR 3.5.1.4 and 3.5.2.5. The requested changes do not serve as initiators of any Columbia accident previously evaluated.

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

The accident probabilities are unaffected and the consequences remain unchanged.

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

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 16 of 19 2) Does the proposed amendment create the possibility of a new or different kind of accident 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 previous evaluations.

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

3) Does the proposed amendment involve a significant reduction in a margin of safety?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 by 5% and 10%, respectively, to ensure no margin of safety was impacted.

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

Therefore, the proposed change does not involve a significant reduction in the margin of safety.Based on the above, Energy Northwest concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

7.0 CONCLUSION

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

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 17 of 19 8.0 ENVIRONMENTAL CONSIDERATION Energy Northwest has determined that the proposed amendment would change requirements with respect to installation or use of a facility component located within Columbia's restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

Energy Northwest has evaluated the proposed change and has 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 any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed change meets the 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 or environmental assessment need be prepared in connection with the proposed amendment.

9.0 REFERENCES

1. Letter G02-93-180, JV Parrish (Washington Public Power Supply System) to NRC, WNP-2 Operating License NPF-21 Request for Amendment to the Facility Operating License and Technical Specifications to Increase Licensed Power Level From 3323 MWt to 3486 MWt With Extended Load Line Limit and Change in Safety Relief Valve Setpoint Tolerance, dated July 9, 1993.2. Letter, JW Clifford (NRC) to JV Parrish (Washington Public Power Supply System), Issuance of Amendment for the Washington Public Power Supply System 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, License Amendment Request for Changes to Technical Specifications Involving Core Operating 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 License Amendment Request Involving Core Operating Limits Report and Scram Time Testing, 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 Scram Time Testing (TAC No. MD9247), dated May 5, 2009.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 18 of 19 7. GE Nuclear Energy, NEDC-32141 P, Power Uprate with Extended Load Line Limit Safety Analysis for WNP-2, June 1993.8. GE Nuclear Energy, GE-NE-208-17-0993, WNP-2 Power Uprate Project NSSS Engineering Report, Revision 1, December 1994.9. Energy Northwest, Columbia Generating Station, Final Safety Analysis Report Amendment 61.10. General Electric (GE) Safety Communication (SC) 06-01 Worst Single Failure for Suppression Pool Temperature Analysis, January 19, 2006.11. GE Hitachi Nuclear Energy, NEDC-33813P, Technical Specification Change Support 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 Change Support 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 at Reduced 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, License Amendment Request to Change Technical Specifications in support of PRNM /ARTS/MELLLA Implementation, dated January 31, 2012.

LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 19 of 19 20. Letter, CO Thomas (NRC) to JF Quirk (GE), Acceptance for Referencing of Licensing 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 Program and 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 CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure I Page 1 of 1 Attachment I Proposed Columbia Technical Specification Changes (Mark-Up)

ECCS -Operating 3.5.1 SURVEILLANCE 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 reactor steam dome pressure less than 48 psig in MODE 3, if capable of being manually realigned and not otherwise inoperable.

Verify each ECCS injectiornspray subsystem manual, power operated, and automatic valve In the flow path, that is not locked, sealed, or otherwise secured in position, Is in the correct position.31 days SR 3.5.1.3 Verify ADS accumulator backup compressed gas 31 days system average pressure In the required bottles is 22200 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 Program DIFFERENTIAL PRESSURE BETWEEN REACTOR AND SYSTEM FLOW RATE SUCTION SOURCE LPCS > gpm > 128 psid LPCI > 746F-7200 gpm > 26 psidg HPCS > 6350 gpm ! 200 psidg SR 3.5.1.5 NOTE ..Vessel injection/spray may be excluded.Verify each ECCS injection/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal.Columbia Generating Station 3.5.1-4 Amendment No. 460,246 225 ECCS -Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify, for each required ECCS injection/spray 31 days subsystem, the piping is filled with water from the pump discharge valve to the Injection valve.SR 3.5.2.4 ---NOTE One low pressure coolant injection (LPCI)subsystem may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned and not otherwise Inoperable.

Verify each required ECCS Injection/spray 31 days subsystem manual, power operated, and automatic valve in the flow path, that Is not locked, sealed, or otherwise secured in position, is in the correct position.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 Program DIFFERENTIAL PRESSURE BETWEEN REACTOR AND-F_.MEO.,RATE SUCTION SOURCE LPCS > 36.0-6200 gpm 128 psid LPCI > 7 4rag7200 gpm > 26 psid§HPCS > 6350 gpm > 200 psidg SR 3.5.2.6 -----NOTE-Vessel injection/spray may be excluded.Verify each required ECCS injection/spray 24 months subsystem actuates on an actual or simulated automatic initiation signal.Columbia Generating Station 3.5.2-3 Amendment No. 414,206 225 LICENSE AMENDMENT REQUEST FOR CHANGE TO EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS Enclosure 1 Page 1 of I Attachment 2 Proposed Columbia Technical Specification Changes (Re-Typed)

ECCS -Operating 3.5.1 SURVEILLANCE 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 reactor steam dome pressure less than 48 psig in MODE 3, if capable of being manually realigned and not otherwise inoperable.

Verify each ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, Is in the correct position.31 days SR 3.5.1.3 Verify ADS accumulator backup compressed gas 31 days system 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 Program DIFFERENTIAL PRESSURE BETWEEN REACTOR AND SYSTEM FLOW RATE SUCTION SOURCE LPCS 6200 gpm > 128 psid LPCI 7200 gpm > 26 psid HPCS _> 6350 gpm 200 psid SR 3.5.1.5 Vessel injection/spray may be excluded.Verify each ECCS injection/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal.Columbia Generating Station 3.5.1-4 Amendment No. 460,2-0 225 ECCS -Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify, for each required ECCS Injection/spray 31 days subsystem, the piping is filled with water from the pump discharge valve to the Injection valve.SR 3.5.2.4 .......-NOTE One low pressure coolant injection (LPCI)subsystem may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned and not otherwise inoperable.

Verify each required ECCS injection/spray 31 days subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.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 Program DIFFERENTIAL PRESSURE BETWEEN REACTOR AND SYSIEM FLOWM RIE SUCTION SOURCE LPCS >6200 gpm _ 128 psid LPCI >7200 gpm _ 26 psid HPCS >6350 gpm >200 psid SR 3.5.2.6 NOTE--Vessel injection/spray may be excluded.Verify each required ECCS Injection/spray 24 months subsystem actuates on an actual or simulated automatic Initiation signal.Columbia Generating Station 3.5.2-3 Amendment No. 469,205 225