ML082610049

Issuance of Amendment No. 210 Revised TS 3.3.5.2 RCIC System Instrumentation, and TS 3.5.2 ECCS-Shotdown to Increase Storage Tank Level

ML082610049
Person / Time
Site:	Columbia
Issue date:	09/30/2008
From:	Lyon C Plant Licensing Branch IV
To:	Parrish J Energy Northwest
Lyon, C F, NRR/DLPM, 415-2296
Shared Package
ML082610046	List: ML082610049 ML082610056
References
TAC MD6176
Download: ML082610049 (13)

Text

September 30, 2008 Mr. J. V. Parrish Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)

Richland, WA 99352-0968

SUBJECT:

COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT RE:

TECHNICAL SPECIFICATION CHANGES ASSOCIATED WITH CONDENSATE STORAGE TANK LEVEL (TAC NO. MD6176)

Dear Mr. Parrish:

The U.S. Nuclear Regulatory Commission (NRC, Commission) has issued the enclosed Amendment No.210 to Facility Operating License No. NPF-21 for the Columbia Generating Station. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated July 26, 2007, as superseded by application dated August 8, 2007, and as supplemented by letters dated November 19, 2007, and June 5 and July 21, 2008.

The amendment revises the requirements of Technical Specification (TS) 3.3.5.2, Reactor Core Isolation Cooling (RCIC) System Instrumentation, and TS 3.5.2, ECCS [Emergency Core Cooling System]-Shutdown, to increase the Condensate Storage Tank level.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

/RA/

Carl F. Lyon, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397

Enclosures:

1. Amendment No. 210 to NPF-21

2. Safety Evaluation cc w/encls: See next page

Pkg. ML082610046; Amendment ML082610049,License/TS Pgs ML082610056

• • previously concurred *memo dated OFFICE LPL4/PM LPL4/LA SRXB/BC SBPB/BC EICB/BC ITSB/BC OGC NAME FLyon GLappert(**) GCranston(*) DHarrison(*) WKemper(*) RElliott ASilvia DATE 9/19/08 9/17/08 7/18/08 8/20/08 9/5/08 9/22/08 9/29/08 OFFICE LPL4/BC LPL4/PM NAME MMarkley FLyon DATE 9/30/08 9/30/08 Columbia Generating Station (6/10/2008) cc:

Chairman Energy Facility Site Evaluation Council Senior Resident Inspector P.O. Box 43172 U.S. Nuclear Regulatory Commission Olympia, WA 98504-3172 P.O. Box 69 Richland, WA 99352-0069 Mr. Gregory V. Cullen Manager, Regulatory Programs Mr. Ken Niles Energy Northwest Assistant Director P.O. Box 968, Mail Drop PE20 Oregon Department of Energy Richland, WA 99352-0968 625 Marion Street NE Salem, OR 97301-3737 Chairman Benton County Board of Commissioners Special Hazards Program Manager P.O. Box 190 Washington Emergency Management Div.

Prosser, WA 99350-0190 127 W. Clark Street Pasco, WA 99301 Mr. William A. Horin, Esq.

Winston & Strawn 1700 K Street, N.W.

Washington, DC 20006-3817 Mr. Matt Steuerwalt Executive Policy Division Office of the Governor P.O. Box 43113 Olympia, WA 98504-3113 Ms. Lynn Albin Washington State Department of Health P.O. Box 7827 Olympia, WA 98504-7827 Technical Services Branch Chief FEMA Region X 130 - 228th Street, SW Bothell, WA 98021-9796 Mr. Mike Hammond Department of Homeland Security FEMA/REP 130 - 228th Street SW Bothell, WA 98021-9796 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125

ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 210 License No. NPF-21

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Energy Northwest (licensee), dated July 26, 2007, as superseded by letter dated August 8, 2007, and as supplemented by letters dated November 19, 2007, and June 5 and July 21, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-21 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 210 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License No. NPF-21 and Technical Specifications Date of Issuance: September 30, 2008

ATTACHMENT TO LICENSE AMENDMENT NO.210 FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of the Facility Operating License No. NPF-21 and Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Facility Operating License REMOVE INSERT Technical Specification REMOVE INSERT 3.3.5.2-4 3.3.5.2-4 3.5.2-3 3.5.2-3

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.210 TO FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397

1.0 INTRODUCTION

By application dated July 26, 2007 to the U.S. Nuclear Regulatory Commission (NRC)

(Agencywide Documents Access and Management System (ADAMS) Accession No. ML072190639), as superseded by application dated August 8, 2007 ADAMS Accession No. ML072330087), and as supplemented by letters dated November 19, 2007 (ADAMS Accession No. ML073381134), and June 5 (ADAMS Accession No. ML081640330) and July 21, 2008 (ADAMS Accession No. ML082110416), Energy Northwest (the licensee) requested changes to the Technical Specifications (TSs) for the Columbia Generating Station (CGS). The requested changes revise the requirements of Technical Specification (TS) 3.3.5.2, Reactor Core Isolation Cooling (RCIC) System Instrumentation, and TS 3.5.2, ECCS [Emergency Core Cooling System]-Shutdown, to increase the Condensate Storage Tank (CST) level.

Specifically, the licensee proposes the following changes:

1. Revise the allowable value for CST Level - Low in Table 3.3.5.2-1, "Reactor Core Isolation Cooling System Instrumentation," from > 446 ft 0 inches elevation to >

447 ft 7 inches elevation

2. Revise the required CST water level in Surveillance Requirement (SR) 3.5.2.2 from

> 14.8 ft in a single CST or > 9.1 ft in each CST to > 16.5 ft in a single CST or >

10.5 ft in each CST The licensee proposed the changes to reflect corrections made by the licensee to the plant-specific supporting calculations for these TS values during its process of documenting the design basis for the RCIC system.

The supplements dated November 19, 2007, and June 5 and July 21, 2008, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 28, 2007 (72 FR 49572).

2.0 BACKGROUND

The condensate supply system is described in the CGS Final Safety Analysis Report (FSAR),

Section 9.2.6. The system is designed to store and provide a condensate supply to the RCIC system, the high pressure core spray (HPCS) system, and the residual heat removal (RHR) loops. The system includes two storage tanks each with a capacity of 400,000 gallons. A minimum inventory of 135,000 gallons in the condensate storage tanks (CSTs) is reserved for the RCIC and HPCS pumps. This ensures the immediate availability of a sufficient quantity of condensate for emergency core cooling, reactor shutdown, and station blackout.

Although a minimum of 135,000 gallons is maintained in the CSTs as a source of water for the RCIC and HPCS pumps, the supply of water in the suppression pool is the emergency source of water for these pumps. This reserve is maintained by monitoring the level in the CSTs and by preventing condensate transfer from the CSTs when this reserve level is reached.

The condensate storage facilities are not required to ensure any of the following:

• The integrity of the reactor coolant pressure boundary,

• The ability to shut down the reactor and maintain it in a safe shutdown condition, or

• The ability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures in excess of the guideline exposure of 10 CFR Part 100.

3.0 REGULATORY EVALUATION

The NRC staff used the following guidance and regulatory requirements in its evaluation of the proposed changes.

In Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, Technical specifications, the Commission established its regulatory requirements related to the contents of the TSs. Specifically, 10 CFR 50.36(d)(1)(ii)(A) states, Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. 10 CFR 50.36(d)(2)(i) states that "Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. In addition, 10 CFR 50.36(d)(3) states, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

10 CFR Part 50, Appendix A, General Design Criterion (GDC) 13, "Instrumentation and control,"

states that "Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and

systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges."

10 CFR Part 50, Appendix A, GDC 20, Protection system functions, states, in part, "The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences...."

10 CFR Part 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, states that "Measures shall be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within necessary limits."

Regulatory Guide (RG) 1.105, Revision 3, Setpoints for Safety-Related Instrumentations, issued December 1999, describes a method acceptable to the NRC staff for complying with the agencys regulations for ensuring that setpoints for safety-related instrumentation are initially within and remain within the TS limits. The RG endorses Part I of Instrumentation, Systems, and Automation Society (ISA)-S67.04-1994, Setpoints for Nuclear Safety Instrumentation, subject to the NRC staff clarifications. The NRC staff used this guide to evaluate the adequacy of the CGS setpoint calculation methodologies and the related plant surveillance procedures.

NRC Regulatory Issue Summary (RIS) 2006-17, NRC Staff Position on the Requirements of 10 CFR 50.36, Technical Specifications, Regarding Limiting Safety System Settings during Periodic Testing and Calibration of Instrument Channels, dated August 24, 2006 (ADAMS Accession No. ML051810077), addresses the 10 CFR 50.36 requirements on limiting safety system settings (LSSSs) assessed during testing and calibration of instrumentation. This RIS discusses why compliance with the allowable values (AVs) in the TSs during testing or calibration alone is not sufficient to ensure that the safety limits (SLs) will be protected until the next periodic surveillance. RIS 2006-17 also suggests (1) verifying that the change in the measured trip setpoint (TSP) during testing or calibration is within predefined limits (acceptable as-found and as-left tolerances), and (2) taking appropriate actions if the TSP is outside these limits, as a method that meets the requirements of 10 CFR 50.36. However, RIS 2006-17 recognizes that other methods and approaches may also be acceptable. The NRC staff used the guidance of RIS 2006-17 to evaluate the effects of the proposed TS change, the acceptability of the setpoint calculation methodology, and the adequacy of the proposed TS change to meet the requirements of 10 CFR 50.36.

The letter from P. Hiland (NRC) to NEI Setpoint Methods Task Force, "Technical Specification for Addressing Issues Related to Setpoint Allowable Values," dated September 7, 2005 (ADAMS Accession No. ML052500004) is complemented by RIS 2006-17. The NRC staff used this letter as guidance to evaluate the adequacy of the proposed TS change to meet the requirements of 10 CFR 50.36.

Revising the TSs to the ranges specified in this submittal will result in the prescribed CST operating ranges being maintained for normal operation, anticipated operational occurrences, transients, and accident conditions. In addition, the establishment of the new ranges will provide the necessary limits for the instruments currently being used to satisfy the affected surveillance

requirements. Therefore, the NRC staff concludes that the proposed amendment complies with the regulatory requirements.

4.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed changes for effects in the areas of reactor systems, balance-of-plant, and instrumentation and controls.

4.1 Proposed Change to TS Table 3.3.5.2-1 TS Table 3.3.5.2-1 provides a list RCIC system functions for which instrumentation is required to be provided. The instruments required to meet function three, Condensate Storage Tank Level-Low are RCIC-LS-15A and 15B. These instruments initiate an automatic transfer of RCIC pump suction from the CST to the suppression pool on low CST level. They are required to be set high enough to ensure adequate pump suction head while water is being taken from the CST. The value for automatic transfer is set high enough to ensure that during swap-over of suction from the CST to the suppression pool, the volume in the CST will remain above the required submergence level to prevent vortexing. The current TS allowable value is > 446 feet, 0 inches.

The licensee discovered that the plant specific calculation (E/l-02-93-1272) used to support the existing values for the CST level setpoint incorrectly based the vortex limit on the centerline of the suction pipe (445 feet, 4 inches) feeding the RClC pump when it should have been based on the upper edge of the pipe (446 feet, 3.625 inches). In order to correct this error, the new lower analytical limit was increased proportionately to 447 feet, 1 inch. Accounting for drift and loop uncertainties at approximately 1 inch, the new calculated minimum setting would need to be 447 feet, 2 inches. To provide additional conservatism, the licensee proposed to add another 5 inches, for a TS allowable value of > 447 feet, 7 inches. Currently, the level switch setting for both level switches are set conservatively at 448 feet, 3 inches in compliance with Licensee Controlled Specifications (LCS) Table 1.3.5.3-1.

The licensee used the square root of the sum of the squares of all independent variables and the algebraic sum of all dependent variables in evaluating the total loop uncertainty (U), total calibration and drift uncertainty (CD), nominal trip setpoint (NTSP), AV, acceptable as-found tolerances, and acceptable as-left tolerances. The licensee calculated U to be one inch and CD to be 0.75 inch. The licensee calculated the AV from the lower analytical limit of 447 feet, 1.00 inch by adding U, subtracting CD, and adding a margin of 5.75 inches. The licensee used 448 feet, 3 inches as NTSP, which provides a reasonable margin from upper analytical limit of 448 feet, 5 inches. The licensee used setting tolerances of +0.50 inches / -0.00 inch. With these numbers, the licensee derived a lower as-left setpoint limit (acceptable lower as-left value) of 448 feet, 3 inches; an upper as-left setpoint limit (acceptable upper as-left value) of 448 feet, 3.50 inches; a lower as-found setpoint limit (acceptable lower as-found value) of 448 feet, 2.25 inches; and an upper as-found setpoint limit (acceptable upper as-found) of 448 feet, 4.25 inches. The NRC staff verified that the licensee specified these as-left and as-found setpoint limits in its plant surveillance procedure ISP-RCIC-Q902, RCIC Suction Transfer on CST Low-Level - CFT/CC, Revision 5.

In its letter dated November 19, 2007, the licensee stated, In the performance of these procedures, "as-found" values that are discovered outside the Technical Specification allowed value are procedurally required to be reported to Operations and entered into Columbia's corrective action program. This results in the performance of an immediate operability determination and actions to restore the condition. Other discovered degraded or non-conforming conditions are similarly addressed through Columbia's corrective action program.

Additionally, if the "as-found" values are discovered outside of the acceptable range for "as-left" settings, the procedure directs re-calibration of the level switch to within the "as-left" limits. The NRC staff finds that CGS procedures are consistent with the guidance of RIS 2006-17 and RG 1.105.

In its application, the licensee stated that the RCIC system is not an Engineered Safety Feature System (ESF) and no credit is taken in the safety analyses for RCIC system operation. The NRC staff reviewed CGS FSAR Section 5.4.6, "Reactor Core Isolation Cooling System," and confirmed that the RCIC system is neither an ECCS nor an ESF system, and no credit is taken in the accident analysis of FSAR Chapter 6 or 15 for its operation. The NRC staff, therefore, concludes that the swap-over function of the RCIC system from the CST to the suppression pool for which a change in the TS allowable value has been proposed is not credited in any safety analysis at CGS. Consequently, the proposed TS allowable value change for CST is not safety limit related.

The NRC staff also concludes that the proposed change to modify the subject TS allowable value to reflect the vortex limit calculated based on the upper edge of the suction pipe feeding the RCIC pump, instead of the centerline of the pipe, is acceptable, since it appropriately corrects a non-conservative value in the TS and is consistent with the guidance of RIS 2006-17 and RG 1.105.

4.2 Proposed Change to SR 3.5.2.2 SR 3.5.2.2 requires that, for operability of the HPCS System, either the suppression pool or the CSTs provide adequate volume (135,000 gallons) to minimize the consequences of a drain down event while in Mode 4 or Mode 5 (if required). There are two CSTs, each with a nominal capacity of 400,000 gallons. For the CSTs to satisfy this surveillance requirement there must be either two tanks, each with 67,500 gallons available, or one tank with 135,000 gallons available.

While in standby, the HPCS pump is normally aligned to the CSTs to minimize the injection of suppression pool water into the Reactor Pressure Vessel (RPV). However, on low CST level or high suppression pool level, an automatic transfer of the HPCS pump suction to the suppression pool occurs upon activation of the level switches. The licensee considered the effects of water flow and measurement inaccuracy in establishing the set point for the CST low level switches (HPCS-LS-1A and 1B). When the level in the CSTs fall below a preset level, contact closure from HPCS-LS-1A and 1B, with one out of two logic, signals the HPCS suction valve from the suppression pool to open. Opening the suppression pool suction valve signals the CST suction valve to close.

The licensee discovered that the calculated minimum level to assure 135,000 gallons in the CSTs did not take into consideration the plant conditions at which automatic transfer from the CST to the suppression pool could take place. These level switches are located on a standpipe

connected to the suction line in the reactor building rather than directly to the CST. Thus, they are influenced by such factors as differential head between the tank and the standpipe (due primarily to flow losses), reactor building temperature, and reactor building pressure. Previously, the calculation did not address the head loss that occurs in the switches when the HPCS pump is drawing water from the CST. When taking into consideration the effects of full HPCS flow on the sensing instrument, it was discovered that premature switching could occur due to a sensed lower level by HPCS-LS-1A and 1B than that which actually existed in the tank.

After taking into consideration the effect that flow had on the level sensors, the licensee determined that an additional margin must be added to the surveillance value to assure 135,000 gallons would be available to the HPCS and RCIC pumps to inject into the RPV prior to reaching the set point for switching suction from the CST to the suppression pool. The licensee determined the additional margin to be 1.4 feet if the water is divided between two CSTs and 1.7 feet if all the water is in one CST.

The licensee described a number of conservative assumptions used in determining the new surveillance level:

• Level switches are assumed to activate at the highest possible point when determining the required level for ensuring 135,000 gallons.

• The ambient air pressure difference between the water in the CST and the standpipe in the reactor building is compensated for by adding 0.6 feet for the two tank elevation and 0.9 feet for the single tank elevation.

• A friction factor of 120% of the Crane Flow of Fluids value is used for the friction factor.

• The analysis does not take into account the additional water available (approximately 1.8 inches, depending on flow rate) after the switch activates and prior to the suction valve switchover taking place.

The NRC staff finds that the conservative assumptions used in the licensees calculation for the new values for SR 3.5.2.2 ensure the minimum volume of 135,000 gallons would be available to the HPCS and RCIC pumps prior to switchover to the suppression pool. Therefore, the proposed values for SR 3.5.2.2 are acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Washington State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, and changes an inspection or surveillance requirement. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or

cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on August 28, 2007 (72 FR 49572). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (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 Commission's regulations, 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.

Principal Contributors: M. Razzaque E. Smith S. Mazumdar Date: September 30, 2008