ML14042A465

From kanterella
Jump to navigation Jump to search

Issuance of Amendments Regarding Water Level in Plant Service Water Pump Well
ML14042A465
Person / Time
Site: Hatch  Southern Nuclear icon.png
Issue date: 05/13/2014
From: Martin R
Plant Licensing Branch II
To: Pierce C
Southern Nuclear Operating Co
Martin, Robert NRR/DORL 415-1493
References
TAC ME9004, TAC ME9005
Download: ML14042A465 (25)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 13, 2014 Mr. C. R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

Post Office Box 1295, Bin - 038 Birmingham, AL 35201-1295

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING WATER LEVEL IN PLANT SERVICE WATER PUMP WELL (TAG NOS. ME9004 AND ME9005)

Dear Mr. Pierce:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 267 to Renewed Facility Operating License DPR-57 and Amendment No. 211 to Renewed Facility Operating License NPF-5 for the Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2, respectively.

The amendments revise the Technical Specifications (TSs) in response to your application dated July 5, 2012, as supplemented by letters dated November 13, and November 30, 2012 and February 22, 2013. -

These amendments revise the TSs Limiting Condition for Operation for the plant service water (PSW) and ultimate heat sink. Specifically, the surveillance requirement for the minimum water level in each PSW pump well of the intake structure would be revised from a value of 60.7 feet mean sea level (MSL) to a value of 60.5 MSL. Your application stated that the revised minimum water level will provide additional operational flexibility during periods of low river levels.

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

Sincerely,

/ ) & ;7'n #--__/)1~

?/~n.~enior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-321 and 50-366

Enclosures:

1. Amendment No. 267 to DPR-57
2. Amendment No. 211 to NPF-5
3. Safety Evaluation cc w/encls: Distribution via Listserv

(

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 267 Renewed License No. DPR-57

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

A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 1 (the facility) Renewed Facility Operating License No. DPR-57 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated July 5, 2012, as supplemented by letters dated November 13, and November 30, 2012 and February 22, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Oper~ting License No. DPR-57 is hereby amended to read as follows: *

(2) Technical Specifications r

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 267, are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This 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 Jl!P~-

Robert Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-57 and the Technical Specifications Date of Issuance: May 13 , 2 0 14

ATTACHMENT TO LICENSE AMENDMENT NO. 267 RENEWED FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License DPR 57, Page 4 DPR-57, Page 4 TSs TSs 3.7-5 3.7-5

for sample analysis or instrument calibration, or associated with radioactive apparatus or components; (6) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, *but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Coi'Jlmission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power levels not in excess of 2804 megawatts thermal.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 26 7 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirement (SR) contained-in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated before the time and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled performance.

(3) Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report for the facility, as contained in the updated Fire Hazards Analysis and Fire Protection Program for the Edwin I. Hatch Nuclear Plant, Units 1 and 2, which was originally submitted by letter dated July 22, 1986. Southern Nuclear may make changes to the fire protection program without prior Commission approval only If the changes Renewed License No. DPR-57 Amendment N0.267

PSW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in each PSW pump well of In accordance with the intake structure is I! 60.5 ft mean sea level the Surveillance (MSL). Frequency Control Program 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when water level is

S 61.7 ft MSL SR 3.7.2.2 ---------------------NOTE--------------------------

Isolation of flow to individual components or systems does not render PSW System inoperable.

Verify each PSW subsystem manual, power In accordance with operated, and automatic valve in the flow paths the Surveillance servicing safety related systems or components, Frequency Control that is not locked, sealed, or otherwise secured in Program position, is in the correct position.

SR 3.7.2.3 Verify each PSW subsystem actuates on an actual ln accordance with or simulated initiation signal. the Surveillance Frequency Control Program HATCH UNIT 1 3.7-5 Amendment No. 26 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 211 Renewed License No. NPF-5

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

A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 2 (the facility) Renewed Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated July 5, 2012, as supplemented by letters dated November 13, and November 30, 2012 and February 22, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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 hereby amended by page changes to the Technical Specifications as indicated in the" attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-5 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 211 are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This 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 Robert Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-5 and the Technical Specifications Date of Issuance: May 13 , 2 014

ATTACHMENT TO LICENSE AMENDMENT NO. 211 RENEWED FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages

. License License NPF-5, Page 4 NPF-5, Page 4 TSs. TSs 3.7-5 3.7-5

(6) Southern Nuclear, pursuimt to the Act and 10 CFR Parts~30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain, and is subject to, the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions 2 specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power levels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 211.

are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shali be made by an amendment to the license supported by a favorable evaluation by the Commission.

(a) Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report for the facility, as contained 2

The original licensee authorized to possess, use, and operate the facility was Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.

Renewed License No. NPF*5 Amendment No.211

PSW System and UHS 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level in each PSW pump well of In accordance with the intake structure is ~ 60.5 ft mean sea level the Surveillance (MSL). Frequency Control Program 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when water level is s 61.7 ft MSL SR 3.7.2.2 -------------------------NOTE-------------------------

Isolation of flow to individual components or systems does not render PSW System inoperable.

Verify each PSW subsystem manual, power In accordance with operated, and automatic valve in the flow paths the Surveillance servicing safety related systems or components, Frequency Control that is not locked, sealed, or otherwise secured in Program position, is in the correct position.

SR 3.7.2.3 Verify each PSW subsystem actuates on an actual In accordance with or simulated initiation signal. the Surveillance Frequency Control Program HATCH UNIT2 3.7-5 Amendment No. 211 I

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 267 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-57 AMENDMENT NO. 211 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-5 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-321 AND 50-366 1.0. INTRODUCTION By license amendment request (LAR) dated July 5, 2012, as supplemented on November 13, and November 30, 2012 and February 22, 2013 (Agencywide Documents Access and Management System (ADAMS), Accession Nos. ML12188A721, ML12319A054, ML12338A029 and ML13053A493, respectively, Southern Nuclear Operating Company, Inc. (SNC, the licensee),

requested changes to the Technical Specification (TS) Surveillance Requirements (SR) for the Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2 (HNP).1 The LAR would revise the minimum water level referenced in the SR 3.7.2.1 associated with the Limiting Condition for Operation (LCO) for the pla~t service water (PSW) system and ultimate heat sink (UHS) (LCO 3.7.2). SNC stated that the revised minimum water level proposed by the LAR would provide additional operational flexibility during periods of low river levels.

The supplements dated November 13 and November 30, 2012, and February 22, 2013, 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 September 4, 2012 (77 FR 53930).

2.0 REGULATORY EVALUATION

2.1 *Balance of Plant The UHS (Aitamaha River) provides the source of cooling water for the PSW System, Residual Heat Removal Service Water (RHRSW) System and the Standby Service Water (SSW) System.

The pumps for each system take suction from the UHS at the intake structure.

1 As noted in the July 5, 2012 LAR, its subject had been addressed by previous SNC submittals. The July 5, 2012 LAR, as supplemented, superceded ir-1 their entirety the previous submittals.

The safety design basis of the PSW system is to provide a reliable' source of cooling water to equipment required for accident conditions. The PSW system also provides cooling water for normal operating and shutdown conditions. The PSW provides cooling water to:

1) turbine building heat exchangers associated with power conversion systems located in the turbine building, 2) reactor building closed cooling water (RBCCW) system heat exchangers,
3) radwaste building closed cooling water heat exchangers, 4) standby diesel generator (SDG) heat exchangers, 5) primary containment (drywell) chiller condensers, 6) residual heat removal (RHR) pump coolers, 7) low pressure coolant injection (LPCI) inverter room coolers, 8) safeguard equipment heating, ventilation, and air-conditioning (HVAC) system, and 9) main control room environmental control (MCREC) system.

The RHR$W system provides cooling water to the RHR heat<exchangers, which are required for a safe reactor shutdown following a design-basis accident. The RHRSW system is operated whenever the RHR heat exchangers are required in shutdown cooling or suppression pool cooling or spray mode of the RHR system.

The SSW for diesel generator (DG) 1B is designed to provide 'cooling water for the removal of heat from the DG 1B, which is the only component served by the SSW system ..

Currently, Surveillance Requirement 3.7.2.1 requires the licensee to verify the water level in each PSW pump well to be:::: 60.7 ft Mean Sea Level (MSL).

The minimum water level is based on providing adequate submergence *of the PSW pumps in order to provide sufficient net positive suction head and to prevent vortexing.

The U. S. Nuclear Regulatory Commission (NRC) requirements, review guidelines, and licensing basis that the staff considered applicable to the License Amendment Request (LAR) include:

  • General Design Criterion (GDC)-44, "Cooling water," requires in part that a system to transfer heat from Structure, System, and Components (SSC) important to safety to an UHS shall be provided. The system safety function shall be to transfer the combined heat load of these SSCs under normal operating and accident conditions.
  • General Design Criterion 5, "Sharing of Structures, Systems, and Components," requires that SSCs important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cool down of the remaining units.

Specifically, the UHS serving multiple units should be capable of providing sufficient cooling water to permit simultaneous safe shutdown and cool down of all units it serves and to maintain them in a safe shutdown condition. Also, in the event of an accident in one

'unit, the UHS should be able to dissipate the heat for that accident safely, to permit the concurrent safe shutdown and cool down of the remaining unit, and to maintain all units in a safe shutdown condition. Sufficient conservatism should be provided to ensure that a

30-day supply of water is available and that the design basis temperatures of safety-related equipment are not exceeded.

2.2 Hydrology Title 10 of the Code of Federal Regulations (1 0 CFR) 50.36, "Technical Specifications," requires that each license authorizing operation of a production or utilization facility include the TS in the application for license. The TS will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. These TS are intended to assure that the plant operates in an analyzed condition. The TS include, among other items, LCOs and SRs. LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the condition can be met. SRs 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 LCOs will be met.

10 CFR 50.36 provides the regulatory requirements for the contents required in a licensee's TS.

Criterion 3 of 10 CFR 50.36(c)(2)(ii) requires an LCO to be established for a structure, system or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The NRC staff's assessment of the proposed LAR, with respect to the proposed change in the minimum water level of the UHS for operability, was performed in accordance with the guidance of the Standard Review Plan (SRP) (NUREG-0800) Section 9.2.5 "Ultimate Heat Sink." This SRP .

section provides guidance for evaluating the capability of water sources for performing the UHS function in accordance with the requirements specified by General Design Criteria (GDC) 2, 5, 44, 45, and 46. The principal GDC establish the necessary design, fabrication, construction, testing, and performance requirements for SSCs important to safety; that is, SSCs that provide reasonable assurance that the 'facility can be operated without undue risk to the health and safety of the public.

3.0 TECHNICAL EVALUATION

3.1 Balance of Plant The UHS current OPERABILITY requirement for the minimum water level in the PSW pump wells is 60.7 feet MSL as specified in Surveillance Requirement (SR) 3.7.2.1. The licensee is proposing to change this OPERABILITY requirement to 60.5 feet MSL by changing SR 3.7.2.1 to require"~ 60.5 feet MSL" instead of"~ 60.7 feet MSL". *The PSW pump wells are part of the intake structure from the Altamaha River. The water level in the PSW pump well is the same level for the RHRSW and SSW pumps since they are located in the common intake structure. The licensee has proposed the revised minimum water level for additional operational flexibility during periods of low river levels.

In order to meet the technical requirements of GDC 44, the PSW pump well minimum level must satisfy the minimum submergence requirements for the PSW, RHRSW and SSW pumps. The NRC staff reviewed the Hatch Updated Final Safety Analysis Report (UFSAR) and pump vendor

drawings to ensure the LAR was in conformance with licensing and existing hydraulic requirements. The minimum submergence requirements from vendor drawings S52586, S56317, S603315, S60778, and SX24108 and are summarized in the table below.

Pump Minimum Water Level for Minimum Submergence (ft MSL)

Unit 1 PSW pumps 59.1 Unit 2 PSW pumps 59.3 Standby Service Water pump 60.0 Units 1 and 2 RHRSW pumps 59.8 The most limiting water level in the intake structure including the PSW pump wells is 60.0 ft MSL.

Therefore, the licensee's proposed minimum level of 60.5 ft MSL satisfies the minimum submergence requirements of the PSW, RHRSW and SSW pumps. The minimum submergence requirements as stated in the table above include both the minimum level for net positive suction head and prevention of vortexing. The components used to monitor water level in the PSW pump well are redundant river level indicators in the main control room and the intake structure and the pump well level stick.

The minimum water level in the PSW pump wells must also be sufficient such that the UHS can perform its safety functions for a minimum of 30 days in order to meet the guidelines of RG 1:27. The safety function includes providing adequate cooling to allow a safe shutdown and cool down of a unit following a design-basis accident and simultaneously provide cooling for safe shutdown and cool down of the other unit.

  • The minimum submergence levels listed by the licensee in the LAR did not completely agree with the vendor drawings. The vendor drawings for the RHRSW pumps list a minimum submergence of 35 inches for a flow rate of 4000 gallons per minute (GPM), whereas the licensee used the same submergence value for a flow rate of 4400 GPM. Similarly the vendor drawing for the SSW pumps lists a minimum submergence of 1.0 foot for a flow rate of 700 GPM, whereas the licensee used the same submergence value for a flow rate of 850 GPM. Therefore, the NRC staff requested additional information in a letter dated January 28, 2013, to explain why the minimum submergence values for these pumps as listed in the LAR were less conservative than the values listed on the vendor drawings. In their response dated February 22, 2013, the licensee stated that they requested the pump vendor provide minimum submergence values for the RHRSW and SSW pumps. The RHRSW pump vendor responded stating that a minimum submergence of 34 inches is required for a flow rate of 4400 GPM and a minimum submergence of 31 inches is necessary for a flow rate of 4000 GPM. The SSW pump vendor stated that 12 inches is the required submergence for either 850 or 700 GPM.

The licensee also obtained updated submergence requirements for 4500 GPM from the PSW pump vendor. Therefore, with the bottom of the intake structure at 56ft MSL and the pump suction bell distance from intake bottom as listed on the vendor drawings, the minimum submergence for the PSW, RHRSW and SSW pumps as listed in the table above are satisfactory with respect to net positive suction head and vortex prevention. The licensee has stated in their letter dated February 22, 2013, that action items have been generated to revise and clarify information on the drawings. The most limiting level is 60.0 feet MSL as provided by vendor drawings and the licensee's correspondence with the pump vendors. Therefore, the licensee's request to change SR 3.7.2.1 to require a water level in each PSW pump well of the intake

structure to be ~ 60.5 feet MSL is acceptable with respect to hydraulic requirements and is in compliance with GDC 5, 44 and RG 1.27.

The licensee's proposal to require the minimum water level in each PSW pump well intake structure to be~ 60.5 feet MSL must also be evaluated with respect to the UHS ability to supply

. the required heat sink for 30 days during a DBA. At 60.5 ft MSL the UHS must last a minimum of 30 days before level reaches 60ft MSL. This issue is evaluated in the following section.

3.2 Hydrology 3.2.1 Licensee's Basis for Minimum Water Level TS change The revised minimum water level in the LAR is based on an updated design basis analysis that

  • revises the minimum river level necessary for the Altamaha River to support post-accident cooling requirements for a 30 day period. The LAR would revise the minimum water level from the current TS value of 60.7 ft MSL to 60.5 ft MSL and therefore would provide additional operational flexibility during periods of low river levels. The licensee performed hydrologic analyses on the river water supply for the UHS to demonstrate that at the minimum water level of 60.5 feet for operation, the HNP UHS function of providing a 30 day cooling water supply would still be met.

The Altamaha River is the HNP UHS and the sole source of water for the PSW, RHRSW and standby service water pumps.

With regards to the hydrologic analysis, the licensee described the methodology used to evaluate the Altamaha River and reach a determination that sufficient water would be available for 30 days for post accident safety cooling. The methodology is described as follows:

River Flow As stated in the LAR, the river discharge rating table at the HNP intake structure was developed by making appropriate adjustments to the discharge data from the nearest United States Geological Survey (USGS) gage, specifically gage 02-2250-00, on the Altamaha River near Baxley, Georgia. The Baxley gage is located on the south bank of the river upstream from the HNP intake structure on the same side. The discharge data at the Baxley gage is developed and periodically updated by the USGS. At the Baxley gage, the USGS performs bathymetric surveys of the river cross-section and measures a river-stage relationship on regular intervals to calculate area. Additionally, velocities are measured at selected locations. The river flow or discharge data at specific elevations at the Baxley gage are developed using the equation Q =VA, where Q =

discharge in cubic feet per second (cfs), V =velocity in feet per second, and A= area in square feet.

In addition, the LAR also states that the USGS revises the discharge rating table at the Baxley gage on an as-needed basis or when additional relevant data is collected, and this data shows the river bottom has stabilized. Presently, the revised USGS rating table is No: 13.1. The river discharge rating table at the HNP intake structure is updated on at least an annual basis per SNC procedure. When updating the HNP intake structure discharge rating table, the latest Baxley gage discharge data is adjusted to:

  • Reflect an estimated static drop of 0.24 ft from the USGS gage to inside the intake structure, specifically a drop of 0.14 ft from the USGS gage to the intake structure and an additional 0.1 ft drop through the traveling water screens
  • Reflect the reference elevation at the Baxley gage of 61.08 feet. The reference elevation is used because a datum elevation is not provided for the current Baxley gage by USGS.

As stated in Reference 4, the PSW pump well minimum water level should be established based on the safe shutdown and post loss-of-coolant accident (LOCA) cooling requirements which consist, per unit, of one PSW pump and two RHRSW pumps. Additionally, there is the separate standby service water pump which provides cooling water for diesel generator "1 B" (shared between Units 1 and 2). Based on consideration of river flow and the limiting minimum submergence of 60.0 ft MSL (based on a minimum submergence of 60.0 ft MSL for the standby service water pump), safe shutdown cooling requirements can be met at a PSW pump well water level of 60.0 ft MSL. Therefore, the proposed PSW pump well minimum water level of 60.5 ft MSL provides margin to assure that the UHS would remain available for 30 days post-LOCA. By interpolating the aforementioned discharge rating table included as Table 2 in the LAR, the river flows would be 718 cfs at 60.5 ft MSL and 517 cfs at 60.0 ft MSL, with both flows sufficient to support safe shutdown cooling requirements.

Thirty Day Supply The licensee used a recession formula, known as a groundwater depletion curve, to determine the number of days it will take for the PSW pump well water level to decrease from the proposed TS SR 3.7.2.1 minimum level of 60.5 ft MSL to the minimum water level at which post-LOCA cooling requirements can be met (60.0 ft MSL).

The recession formula is explained in more detail in Reference 5 and is expressed as Qt = Qoe-oct Where QO is the initial stream flow and Qt is the stream flow at timet. The term e-oct is replaced by "k", called the recession constant or depletion factor, which is commonly used as an indicator of the extent of base flow. The typical ranges of recession constants for streamflow components are 0.2-0.8 for runoff (hi water), 0.7-0.94 for interflow and 0.93-0.995 for groundwater flow (base flow). These values overlap. In general, high recession constants (e.g., > 0.9) tend to indicate dominance of base flow in stream flow (Reference 5). In addition, the "k" value typically used at HNP is 0.995; the range for the "k" value being from 0.995,to 0.997. With the applicable HNP recession constant or "k" value determined, flow values from the HNP intake structure discharge rating table and the recession formula are used to project the time (in days) for the PSW pump well level to reach 60.0 ft MSL when starting from the proposed TS SR 3.7.2.1 minimum level of 60.5 ft MSL.

This projection demonstrates that it would take at least 30 days for this drop in pump well level to occur. Specifically, based on Table 2 of Reference 3, the duration for the water level to drop from the proposed TS SR 3.7.2.1 minimum level of 60.5 ft MSL to 60.0 ft MSL would be 66 days.

Therefore, the licensee concluded that application of the recession formula demonstrates that there will continue to be significant margin to assure that the UHS would remain available for 30 days post-LOCA. Finally, the time (in days), based on the recession formula, for the PSW pump well level to reach 60.0 ft MSL, when starting from the TS SR 3.7.2.1 minimum level, is updated at least annually, per SNC procedure. The river discharge rating table at the HNP intake structure is also updated annually.

Independent Evaluation by USGS In addition to the information provided in the LAR, the licensee also submitted the results of an independent review performed by the USGS. Specifically, the USGS reviewed the methodology used by HNP to determine the Hatch river flow rating calculation of 2008. The purpose of the review was to determine and verify that there was sufficient water supply at the river intake for low flows. In its response letter included in Enclosure 4 of the LAR, the USGS stated that "Our review indicates that your methods are conservative and satisfactory for your stated objective."

Additional Contributing factors Other factors that provide reasonable assurance that the levels in the Altamaha River are predictable are as follows:

  • the drainage basin for the Altamaha River is 11,600 sq. miles in size. Its large size increases the probability of the availability of water.
  • two dams supply the Altamaha River, the Lloyd Shoals dam on the Ocmulgee and the Sinclair dam on the Oconee. Georgia Power Company (GPC) has a Federal Energy Regulatory Commission (FERC) drought contingency plan that states that GPC will release a minimum of 250 cfs from Sinclair: even when inflows are less. At Lloyd Shoals, the minimum release is 400 cfs, or inflow, whichever is less. However, Lloyd Shoals recently has maintained 250 cfs, even when inflows were less. Therefore, the combined minimum discharge from Sinclair and Lloyd Shoals dams is around 500 cfs.
  • the HNP site is underlain by a shallow unconfined aquifer and a deeper lying, minor confined aquifer. The unconfined aquifer lies above elevation 100 to 120 ft MSL in the area of HNP, with the unconfined water table generally reflecting the site topography. The minor confined aquifer consists of silty sands of the Hawthorn Formation between approximate elevations 65 and 0 ft MSL. Piezometric levels generally are below the elevation of 80 ft MSL, and the potentiometric surface slopes northeastward toward the Altamaha River. The Altamaha is hydraulically connected to the two aquifers.

3.2.2 NRC Staff Evaluation This review by the NRC staff focused in the following two primary areas:

a) Adverse effects, if any, on the structural integrity of the SSCs affected by the proposed

  • conditions in the LAR, b) The general hydrologic methodology used to develop the LAR.

The NRC staff notes that the GDC in Appendix A of 10 CFR Part 50 were published after the issuance of the HNP Unit 1 (HNP-U1) construction permit and therefore are only applicable to HNP, Uni.t 2. As such, conformance to the GDC requirements is different for Unit 1 and Unit 2.

For HNP-U1, the design criteria are detailed in the Final Safety Analysis Report (FSAR) Appendix F, Section F.1 and follow the 70 GDC which were proposed in July 1967. A detailed comparison vilas performed by the licensee between the proposed GDC in Appendix F of the HNP-U1 FSAR (to which the facility was licensed) and the current GDC in 10 CFR 50 Appendix A. This analysis is described in HNP-U1 FSAR, Appendix F, Section F.3. Based on a review of this comparison performed by the licensee, the staff determined that the HNP-U1 specific GDC criterion are similar to the current applicable Appendix A GDC criterion 2, 5, 44, 45, and 46, as stated in the

SRP Section 9.2.5 for UHS. Therefore, the NRC staff utilized and reviewed the LAR submittal against the regulatory requirements of the GDC in Appendix A of 10 CFR Part 50, as indicated above.

Structural Integrity of SSCs Affected by the Proposed TS Change The NRC staff evaluated the potential for adverse effects on SSCs important to safety using the guidance described in the SRP Section 9.2.5. This SRP section provides guidance for evaluating the capability of water sources for performing the UHS function in accordance with the requirements specified by GDC 2, 5, 44, 45, and 46. After staff review of the LAR and evaluation of the GDC criterion referenced above, the NRC staff finds revising the river level associated with LCO 3. 7.2 to a lower level results in no adverse effect on the structural integrity of any SSCs because the decreased water level has no impact, within the scope of review of this SE, for each GDC criterion. The staff determined that there is reasonable assurance that the GDC criterion for structural stability applicable to the UHS will continue to be met. Therefore, the NRC staff finds that the proposed change in the LAR is acceptable with regards to the structural integrity of SSCs.

Hydrology Methodology Evaluation Process The revised minimum water level in the proposed LAR is based on an updated design basis analysis that changes the minimum river level necessary for the Altamaha River to support post-accident cooling requirements for a 30 day period. As previously stated, the.scope of this review focused on verifying the adequacy of the methodology used in this updated design basis analysis. Demonstration of appropriate hydrologic methodology may be accomplished by using well known and proven methodologies and procedures with pertinent references, using sound engineering judgment that demonstrates sufficient conservatisms, and by having, whenever possible, a peer review by an independent third party with knowledge on the matter.

This review sought to verify and confirm that there is reasonable assurance that sufficient water is available to support post-accident cooling requirements for a 30 day period and that all required SSCs important to safety function properly, as intended. Demonstration of reasonable assurance that sufficient water is available to support post-accident cooling requirements is accomplished by proving that the PSW pump well water level remains above 60.0 ft MSL for at least 30 days (plus added margin) and therefore the UHS would remain available post-LOCA.

Evaluation of River Flow Analysis With regards to the river flow data obtained by HNP using the methodologies described in Section 3.1 of this SER, the NRC staff evaluated the LAR and determined that the proposed revised conditions in the LAR have no impact on the ability or capability of the HNP to calculate and develop a discharge rating table, based on the following:

  • The process is site-specific and requires the interaction between the USGS and HNP.
  • The general discharge data at the Baxley gage is developed and periodically updated by the USGS and not by HNP.

This process is independent of this LAR and is not affected by the proposed changes by the licensee. For the development of the site-specific HNP intake structure discharge table, the Baxley data is then adjusted to accommodate the site-specific factors described in Section 3.1 of

this safety evaluation (SE). However, this LAR does not change, modify or revise the adjustment factors currently used and approved for HNP. After reviewing the information presented in the LAR, the NRC staff determined that there is no adverse effect in the adjustment process or methodology of the Baxley data used by SNC for the proposed LAR. Therefore, the staff finds that the proposed LAR change is acceptable with regards to the evaluation of river flow analysis.

Evaluation of Thirty Day Supply For the calculation of the thirty day supply using the recession equation described in Section 3.1 of his SE, the NRC staff requested additional information from the licensee by letter dated October 31, 2012 (Reference 6). Specifically, the staff reviewed the technical paper discussed in Reference 5 which stated that the overlapping ranges for the selection of "k" values represent the inherent difficulties in identifying a particular recession as being surface runoff, interflow, or base flow. Furthermore, Reference 5 also listed the range for baseflow as being between 0.93 -0.995 and the LAR states that its baseflow "k" range goes from 0.995 -0.997.

The NRC staff requested the licensee to provide additional information regarding the general process, criteria and analysis that has been used in order to determine the recession constant "k" value of 0.995 for the site, with special emphasis on how the aforementioned limitations of Reference 5 were resolved. These limitations included, but were not limited to, establishing distinctions between interflow and baseflow, seasonally variable components of baseflow analyzed over a specified period of time, etc. By letter dated November 30, 2012 (Reference 7),

the licensee responded that:

A review of river flow data for the last 12 years supports a value of 0.995 for 'k' as demonstrated by the following Figure 1 [in Reference 7]. For each day on the annual timeline the respective maximum flow, median flow, and minimum flow is selected from the river flow data since the year 2000. Figure 1 [in Reference 7] shows this data graphically with the maximum flow line showing the highest of the high flows for each day and the minimum flow line showing the lowest of the low flows for each day for years 2000 through 2011. r In addition, the licensee indicated in its response that by using data from the minimum flow line over a selected period of days, the value of 'k' can be determined by using two points 30 days apart from the minimum flow line on Figure 1 [in Reference?]:

00=1650 cfs--June 4 point from minimum flow line Qt=1400cfs-- July 4 point from minimum flow line t=30 days Qt = Qoe-rx.t 1400=1650

  • k30 1400/1650= k30 Log(1400/1650) =30*Log k Log(1400/1650)/30=Log k 10 Log(1400/1650)/30 = k 0.99454 = k

The licensee also ~tated that the "k" value can be verified using the river flows interpolated from the intake structure discharge table. The intake discharge data was included in the July 5, 2012 LAR, Enclosure 1, Table 2 for 60.5 ft MSL and 60.0 ft MSL, which yielded a river discharge by interpolation of 718 cfs and 517 cfs, respectively.

00=718 cfs 517=718

  • k66 Qt=517cfs Log (517/718)/66 =Log k t=66 days 10Log (517/718)/66 = k 0.995036= k The licensee also responded in Reference 7 that the distinction between interflow and baseflow is determined by a graphical review of the historical river flow data in Figure 1, specifically looking for a change in slope of the minimum flow line. Based on the graphical review, flows above 2000 cfs are generally considered interflow and flows below 2000 cfs are considered baseflow.

However, the distinction of interflow versus baseflow is not relevant to a conservative demonstration of being able to support post-accident cooling requirements for 30 days. Use of a

'k' value appropriately supported by river flow data is technically justified.

Based on this analysis and the additional information provided in the response, the NRC staff determined that the licensee demonstrated that a value of "k" of 0.995 is adequate to use in the recession equation. The value is supported by historical data provided in the Reference 7 response and was verified against data provided in the LAR.

The NRC staff notes that any individual recession is a short-term event that may vary from successive recessions based on several factors. Therefore, in longer period events, it is necessary to combine individual base flow recessions in order to provide an average characterization of base flow, called a master recession curve (MRC). In Reference 6, the staff requested the licensee to provide additional information that would clarify the threshold of what constitutes a short-term event versus a long-term event in terms of having to calculate a single recession or an MRC curve. The staff sought to verify that the option selected to characterize baseflow at HNP, Units 1 and 2, is acceptable in comparison to the single versus average methods and demonstrate that the most frequent depletion simulation has been obtained based on the method selected and, therefore, used to determine the projected time to reach 60.0 ft MSL.

In Reference 7, the licensee's response indicated that the 'k' value used by HNP is based on historical river flow data at the HNP intake from years 2000 through 2011 and appropriate consideration of the trend of current river flow data when operating under HNP low river level procedural actions and that this methodology appears to approximate the use of a MRC curve.

SNC also indicated that use of river flow data from years 2000 through 2011 supports development of a 'k' value based on long-term trends versus. potential short-term perturbations.

As stated previously by SNC, when operating under HNP low river level procedural actions as river level gets closer to the TS SR 3.7.2.1 minimum level, appropriate consideration of the trend of current short-term river flow data may result in a change to the 'k' value. However, based on past experience, SNC expects use of a 'k' value different than 0.995 to be rare.

Based on this analysis and the additional information provided in the response, the NRC staff determined that the licensee demonstrated that the option selected to characterize baseflow at HNP, Units 1 and 2, demonstrates that the most frequent depletion simulation has been obtained

based on the method selected. The values obtained have been supported by historical data provided in the response and verified against data provided in the LAR.

With regards to the proposed PSW pump well minimum water level of 60.5 ft MSL, the NRC staff reviewed the methodology used in order to provide reasonable assurance, with margin, that the proposed level would assure that the UHS would remain available for at least 30 days post-LOCA.

The licensee stated that the recession equation, used with a 'k' value of 0.995, would yield the estimated amount of days that the UHS would be able to supply the required volume of water.

Using the same equation and substituting fort, it is shown that:

Q0=718 cfs 517=718 * .995t Qt=517cfs Log (517/718)= t *Log (0.995) k=0.995 t = -0.143/-.002177 t=65.6 days Based on this analysis and the additional information provided in the response, the NRC staff determined that the licensee demonstrated that the proposed minimum water level of 60.5 ft MSL would still provide the required 30 days post-LOCA cooling requirements. The results also demonstrate that sufficient margin exists to determine that added conservatisms still exist in the proposed conditions. The staff evaluated the information provided in the LAR and the responses to the request for additional information and also determined that there is reasonable assurance that the hydrology method used by HNP to evaluate river flow and determine the availability of water for accident cooling for a 30-day supply is conservative and adequate. Therefore, the NRC staff finds that the proposed change in the LAR is acceptable with regards to the hydrology methodology used in the analysis.

Evaluation of the Independent Review by USGS The NRC staff also verified that the independent evaluation by another Federal agency, the USGS, which was submitted by the licensee in Enclosure 4 of the LAR and described in Section 3.1.4, was appropriate. The staff noted that in this particular case, USGS is an independent entity with expertise on the subject. The USGS reviewed the calculations performed by HNP and determined that the methodology was conservative and satisfactory. Based on these findings, the staff determined that the licensee has demonstrated an adequate review methodology by using well known hydrology methods and procedures that were satisfactorily reviewed by an independent third party with expertise in the subject matter. The NRC staff notes that this satisfactory review of the analysis performed for HNP by another Federal agency reinforces the adequacy of the methodology selected and ensures the proper evaluation to determine that adequate cooling water will be available to meet the TS requirement. Therefore, the NRC staff finds that the proposed change in the LAR is acceptable with regards to the independent third party review performed by USGS.

Evaluation of Additional Contributing factors Based on the discussion in the section on Additional Contributing Factors above, the NRC staff concludes that based on the geographical location and configuration of the Altamaha River and the HNP, there are added features that provide additional conservatism for the UHS analysis

performed in the updated design basis review. Therefore, the staff finds that the consideration of

  • the additional contributing factors is acceptable.

Hydrology Methodology Evaluation Summary The NRC staff reviewed the licensee's analysis provided in the LAR and the responses provided in the RAis. This information was evaluated against the regulations and guidance described in Section 2 of this SE. With regard to the scope of review, the staff determined that there is reasonable assurance that the SSCs associated with the LAR will continue to perform their intended functions during normal and abnormal operating conditions, as stated in the applicable safety analyses documented in the HNP FSAR.

  • Based on this evaluation, the NRC staff determined that there is reasonable assurance that the UHS would remain available for 30 days of emergency cooling, post-LOCA, if the minimum water level of 60.7 ft MSL, as currently stated in plant TS, is revised to 60.5 ft MSL. Therefore, the staff finds that the proposed change is acceptable.

3.2.3 Summary The NRC staff determined that the licensee provided an adequate technical justification which shows that there are no adverse effects on the structural integrity of the SSCs affected by the proposed change. There is reasonable assurance that the SSCs associated with the LAR will continue to perform their intended functions during normal and abnormal operating conditions, as stated in the applicable safety analyses, documented in the HNP FSAR. Additionally, the staff determined that the licensee provided an adequate technical justification that shows that the general hydrologic methodology used to develop the LAR is ac~eptable. Therefore, the NRC staff .

concludes that the licensee's LAR, related to the revised minimum water level referenced in the Units 1 and 2 TS surveillance requirement, associated with the limiting condition for operation for the plant service water system and ultimate heat sink (LCO 3.7.2); is acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on April16, 2014 (79 FR 21487).

Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.

6.0 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) there is reasonable assurance that such activities will be conducted

in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1) Letter from M. J. Ajluni, Southern Nuclear Operating Company, Inc., to NRC Document Control Desk, "Edwin I. Hatch Nuclear Plant, Units 1 and 2, NRC Docket Nos. 50-321 and 50-366- License Amendment Request- Revise the Minimum Water Level in the Plant Service Water Pump Well," dated December 15, 2011. (ADAMS Accession No.:

ML113500108)

2) Letter from M. J. Ajluni, Southern Nuclear Operating Company, Inc., to NRC Document Control Desk, "Edwin I. Hatch Nuclear Plant, Units 1 and 2, NRC Docket Nos. 50-321 and 50-366- Withdrawal of License Amendment Request- Revise the Minimum Water Level in the Plant Service Water Pump Well," dated April 20, 2012. (ADAMS Accession No.:

ML12115A173)

3) Letter from M. J. Ajluni, Southern Nuclear Operating Company, Inc., to NRC Document Control Desk, "Edwin I. Hatch Nuclear Plant, Units 1 and 2, NRC Docket Nos. 50-321 and 50-366...:. License Amendment Request- Revise the Minimum Water Level in the Plant Service Water Pump Well," dated July 5, 2012. (ADAMS Accession No.: ML12188A721)
4) NRC letter to R. P. McDonald, "Issuance of Amendment Nos. 154 and 92 to Facility Operating Licenses DPR-57 -and NPF-5 -Edwin I. Hatch Nuclear Plant, Units 1 and 2 (TACS 62127/62128)," May 12, 1988.
5) R. J. Nathan and T. A. Me Mahan, 1990, "Evaluation of Automated Techniques for Baseflow and Recession Analysis," Water Resources Research 26(7): 1465-1473
6) Letter from R. Martin, USNRC, to M.J.Ajluni, Southern Nuclear Operating Company; "Edwin I. Hatch Nuclear Plant, Units 1 and 2, NRC Docket Nos. 50-321 and 50-366-Request for Additional Information Related to License Amendment Request- Revise the Minimum Water Level in the Plant Service Water Pump Well," dated October 31, 2012.

(ADAMS Accession No.: ML12297A246)

7) Letter from M. J. Ajluni, Southern Nuclear Operating Company, Inc., to NRC Document Control Desk, "Edwin I. Hatch Nuclear Plant, Units 1 and 2, NRC Docket Nos. 50-321 and 50-366- Response to Request for Additional Information to License Amendment Request

- Revise the Minimum Water Level in the Plant Service Water Pump Well," dated November 30, 2012. (ADAMS Accession No.: ML12338A029)

Principal Contributor: Gerard Purciarello Juan Uribe Date: May 13, 2014

ML14042A465 *SE transmitted by memo dated OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/EMCB/BC NRR/DSS/SBPB/BC NAME RMartin SFigueroa AMcMurgtray GCasto DATE 05/13/14 *02/19/14 05/24/13* 05/1 0/13*

OFFICE OGC NRR/LPL2-1 /BC NRR/LPL2-1 /PM NAME A Gosh RPascarelli RMartin DATE 04/28/14 05/12/14 05/13/14

Retrieved from "https://kanterella.com/w/index.php?title=ML14042A465&oldid=2634797"