ML090290209
| ML090290209 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  (NPF-029) |
| Issue date: | 02/25/2009 |
| From: | Lyon C Plant Licensing Branch IV |
| To: | Entergy Operations |
| Lyon C Fred, NRR/DORL/LPL4, 301-415-2296 | |
| References | |
| TAC MD4675 | |
| Download: ML090290209 (18) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 25, 2009 Vice President, Operations Entergy Operations, Inc.
Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150
SUBJECT:
GRAND GULF NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: CONDENSATE STORAGE TANK LOW LEVEL SETPOINT CHANGE (TAC NO. MD4675)
Dear Sir or Madam:
The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 181 to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1. This amendment consists of changes to the Technical Specifications (TSs) in response to the application from Entergy Operations, Inc. (Entergy, the licensee) dated March 1,2007, as supplemented by letters dated September 5 and September 21, 2007, February 14, 2008, and January 19 and February 20, 2009.
The amendment revises the allowable values in the Grand Gulf Nuclear Station, Unit 1, TS Tables 3.3.5.1-1 and 3.3.5.2-1 for the Condensate Storage Tank (CST) low level setpoints for the High Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) suction swap from the CST to the Suppression Pool. The changes correct nonconservative values in the TSs. Currently, the licensee is administratively controlling the nonconservative values in accordance with the guidance of NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety."
The NRC staff issued Regulatory Issue Summary 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," on August 24, 2006.
As discussed in RIS 2006-17, the NRC staff intends to incorporate setpoint issue guidance in the final approved TSTF traveler, which the staff then intends to issue as a consolidated line item improvement process. The NRC staff believes this will establish a uniform, satisfactory resolution that addresses the industry and staff concerns with instrument settings, conforms to the staffs operability guidance, and ensures compliance with Section 50.36 of Title 10 of the Code of Federal RegUlations.
You provided the following regulatory commitment in your letter dated February 20, 2009:
Entergy will follow the efforts of the Technical Specification Task Force (TSTF) and NRC to finalize the details and scope of the changes needed to resolve the instrument setpoint issue discussed in RIS-2006-17. If the Condensate Storage Tank (CST) Level-Low setpoint is affected by the approved TSTF traveler, then Entergy will submit a separate amendment request to implement the approved generic change within 6 months of approval of the TSTF Traveler.
- 2 In its review of the application, the NRC staff had several discussions with the licensee staff regarding whether or not the CST level setpoints are safety limit (SL)-related, in accordance with RIS 2006-17. The licensee's position is that the setpoints associated with the suction swap for HPCS are not SL-related. The NRC staff disagrees, as documented in its letter to the licensee dated November 5,2008. However, since (1) the setpoint methodology and surveillance testing procedures used by the licensee are consistent with the guidance of RIS 2006-17, and the proposed changes will (2) correct an error in the TSs by conservatively increasing the HPCS and RCIC suction swap values, (3) not present an immediate safety concern for current plant operation, and (4) end the prolonged administrative control of the TS values, the NRC staff finds the proposed changes acceptable. Resolution of the control of the CST swapover setpoints within TSs will be addressed pending final approval of the TSTF traveler, as documented in the licensee's letter dated February 20, 2009.
A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, Carl F. Lyon, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-416
Enclosures:
- 1. Amendment No. 181 to NPF-29
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS, INC.
SYSTEM ENERGY RESOURCES, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION ENTERGY MISSISSIPPI, INC.
DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 181 License No. NPF-29
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee), dated March 1, 2007, as supplemented by letters dated September 5 and September 21, 2007, February 14, 2008, and January 19 and February 20, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and 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 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-29 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. 181 are hereby incorporated in the license.
Entergy Operations, Inc. 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 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-29 and the Technical Specifications Date of Issuance: February 25, 2009
ATTACHMENT TO LICENSE AMENDMENT NO. 181 FACILITY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 Replace the following pages of the Facility Operating License No. NPF-29 and the Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Operating License Remove -4 Technical Specifications Remove 3.3-41 3.3-41 3.3-47 3.3-47
(b)
SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above authorized financial transactions, (ii) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS Unit 1 that would materially alter the representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment No. 54.
In addition, SERI is required to notify the NRC of any action by a lessor or other successor in interest to SERI that may have an effect on the operation of the facility.
C.
The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Entergy Operations, Inc. is authorized to operate the facility at reactor core power levels not in excess of 3898 megawatts thermal (100 percent power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 181 are hereby incorporated into this license.
Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
The Surveillance Requirements (SRs) for Diesel Generator 12 contained in the Technical Specifications and listed below, are not required to be performed immediately upon implementation of Amendment No. 169 The SRs listed below shall be successfully demonstrated at the next regularly scheduled performance.
SR 3.8.1. 9, SR 3.8.1.10, and SR 3.8.1.14 Amendment: 181 4
ECCS Instrumentation 3.3.5.1 Table 3.3.5.1-1 (page 3 of 5)
Emergency Core Cooling System Instrumentation APPLICABLE CONDI TJ ONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER REQUIRED SURVEI LLANCE ALLOWABLE FUNCTION CONDITIONS FUNCT I ON ACTION A.l REQUIREMENTS VALUE
- 3.
High Pressure Core Spray (HPCS) System
- a.
Reactor Vessel 1,2,3, 4(b)
~ -43.8 inches Water Level -
Low Low, Level 2 4(a),5(a)
SR SR 3.3.5.1.2 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6
- b.
Drywell 1, 2,3 4(b)
~ 1. 44 psig Pressure Hi gh SR SR 3.3.5.1.2 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6
- c.
Reactor Vessel 1, 2,3, 2
~ 55.7 inches Water Level-High, 4(a),5 (a)
SR SR 3.3.5.1.2 3.3.5.1.3 Level 8 SR 3.3.5.1.5 SR 3.3.5.1.6
- d.
Condensate Storage 1,2,3, 2
~4.7ft Tank Level -
Low 4(c),5 (c)
SR SR 3.3.5.1.2 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6
- e.
Suppression Pool Water Level-High 1,2,3 2
D SR SR 3.3.5.1.1 3.3.5.1.2
~ 7.0 inches SR 3.3.5.1.3 SR 3.3.5.1.5 SR 3.3.5.1.6
- f.
HPCS Pump Discharge Pressure-High (Bypass) 1,2,3, 4(a),5 (a)
E SR SR SR SR 3.3.5.1.1 3.3.5.1.2 3.3.5.1.3 3.3.5.1.5
~
~
108 psig and 1282 psig SR 3.3.5.1.6
- g.
HPCS System Flow Rate-Low (Bypass) 1,2,3, 4(a),5 (a)
E SR SR SR 3.3.5.1.1 3.3.5.1.2 3.3.5.1.3
~
~
1124 1327 gpm gpm and SR 3.3.5.1.5 SR 3.3.5.1.6
- h.
Manual Initiation 1,2,3, C
SR 3.3.5.1.6 NA 4(a),5 (a)
(continued)
(al When associated ECCS subsystem(s) are required to be OPERABLE per LCD 3.5.2, ECCS-Shutdown.
(bl Also required to initiate the associated diesel generator.
(cl When HPCS is OPERABLE for compliance with LCO 3.5.2, "ECCS-Shutdown," and aligned to the condensate storage tank while tank water level is not within the limit of SR 3.5.2.2.
GRAND GULF 3.3-41 Amendment No. -+/--e-9-,
181
RCIC System Instrumentation 3.3.5.2 Table 3.3.5.2-1 (page 1 of 1)
Reactor Core Isolation Cooling System Instrumentation FUNCTION REQUIRED CHANNELS PER FUNCT10N CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- l. Reactor Vessel Water Level - Low Low, Level 2 4
B SR SR SR SR SR 3.3.5.2.1 3.3.5.2.2 3.3.5.2.3 3.3.5.2.4 3.3.5.2.5
? -43.B inches
- 2.
Reactor Vessel Water Level-High, Level 8 2
C SR SR SR SR 3.3.5.2.1 3.3.5.2.2 3.3.5.2.4 3.3.5.2.5
~ 55.7 inches
- 3. Condensate Storage Tank Level - Low 2
D SR SR SR SR 3.3.5.2.1 3.3.5.2.2 3.3.5.2.4 3.3.5.2.5
? 3.7 ft
- 4.
Suppression Level-High Pool Water 2
D SR SR SR SR 3.3.5.2.1 3.3.5.2.2 3.3.5.2.4 3.3.5.2.5
~ 7.0 inches
- 5. Manual Initiation C
SR 3.3.5.2.5 NA GRAND GULF 3.3-47 Amendment No. +&Q.,
181
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 181 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC., ET AL.
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416
1.0 INTRODUCTION
By application dated March 1, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML070670083), as supplemented by letters dated September 5 and September 21, 2007, February 14, 2008, and January 19 and February 20, 2009 (ADAMS Accession Nos. ML072550241, ML072700886, ML080580196, ML090210237 and ML090510684, respectively), Entergy Operations, Inc. (the licensee), requested changes to the Technical Specifications (TSs) for Grand Gulf Nuclear Station, Unit 1 (GGNS). The proposed changes would revise the allowable values (AVs) in the GGNS, TS Tables 3.3.5.1-1 and 3.3.5.2-1 for the Condensate Storage Tank (CST) low level setpoints for the High Pressure Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) suction swap from the CST to the Suppression Pool.
Specifically, the licensee proposes the following changes:
In TS Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation,"
Function 3.d, "High Pressure Core Spray (HPCS) System, Condensate Storage Tank Level-Low," increase the AV from"~ -3 inches" to"~ 4.7 ft."
In TS Table 3.3.5.2-1, "Reactor Core Isolation Cooling System [(RCIC)]
Instrumentation," Function 3, "Condensate Storage Tank Level-Low," increase the AV from "~ -3 inches" to "~ 3.7 ft."
The licensee stated that the proposed TS changes are necessary to correct an error in the original plant design. The licensee's calculation JC-Q1 E22-N654-1, Revision 1, indicates that the error can be resolved by calculating the affected setpoints from the top of the CST suction pipe. Currently, the nonconservative setpoints are being controlled administratively to higher levels in conformance with U.S. Nuclear Regulatory Commission (NRC) Administrative Letter (AL) 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety," dated December 29, 1998.
- 2 The supplements dated September 5 and September 21, 2007, February 14, 2008, and January 19 and February 20, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 8, 2007 (72 FR 26176).
2.0 BACKGROUND
The CST is designed to provide condensate water for the HPCS and RCIC systems during emergency conditions, as well as to serve other functions. Updated Final Safety Analysis Report (UFSAR) Section 7.3.1.1.3.3 states that the CST is the normal source of water supply for the HPCS system and, when the water level in the CST falls below a pre-selected level, HPCS is automatically transferred to the suppression pool. Normally, the suction valves between the HPCS/RCIC and the CST are open, and, upon receiving an HPCS/RCIC initiation signal on the reactor vessel water level falling to the "Reactor Vessel Water Level-Low Low, Level 2," the water is pumped from the CST to the core. When the water level in the CST falls below the "Condensate Storage Tank Level-Low" setpoints, the HPCS and RCIC suctions are automatically transferred to the suppression pool, with the suppression pool suction valves opening before the CST suction valves close.
The CST is the preferred source of water for the HPCS and RCIC system, but the CST is not a seismic Category 1 structure, and, therefore, it is not credited for design-basis accidents (DBAs). Because of this limitation of the CST, the accident analyses credit the suppression pool as the HPCS suction source, which further amplifies the importance of the automatic transfer of the HPCS and RCIC suctions from the CST to the suppression pool upon the "Condensate Storage Tank Level-Low" actuations. In the event of a failure of the non-safety related HPCS/RCIC piping, an automatic transfer from the CST to the suppression pool will occur.
In addition, UFSAR Section 6.3.1.2.1 states, "The primary purpose of HPCS is to maintain reactor vessel inventory after small breaks which do not depressurize the reactor vessel. HPCS also provides spray cooling heat transfer during breaks in which the core is calculated to uncover." HPCS is part of the emergency core cooling system (ECCS) and UFSAR Section 6.3.1.1.1 states that the ECCS is designed to protect any postulated loss-of-coolant accidents (LOCAs) caused by rupture of the primary coolant system piping.
3.0 REGULATORY EVALUATION
The NRC staff reviewed the proposed TS changes in the application against the regulatory requirements and guidance listed below to ensure that there is reasonable assurance that the systems and components affected by the proposed TS changes will perform their safety functions.
3.1 Regulatory Requirements:
3.1.1 In Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.2, "Definitions," safety-related structures, systems and components (SSCs) are defined as those SSCs that are relied upon to remain functional during and following design basis
- 3 events to assure: (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in
§ 50.34(a)( 1) or § 100.11 of this chapter, as applicable.
3.1.2 In 10 CFR 50.36, "Technical specifications," the Commission established its regulatory requirements related to the contents of the TSs. This section states, "Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section."
Furthermore, 10 CFR 50.36(c)(1 )(ii)(A) states, "Limiting safety system settings for nuclear reactors are setting for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting [LSSS] 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 (SL) is exceeded." These LSSS are referred to SL-related LSSS and non SL-related LSSS. See the reference to RIS 2006-17 below.
Specifically, 10 CFR 50.36(c)(2) defines limiting conditions for operation (LCD) as "the lowest functional capability or performance levels of equipment required for safe operation of the facility."
In addition, 10 CFR 50.36(c)(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 of operation will be met." The staff reviewed the proposed TS changes against these 10 CFR 50.36 requirements to ensure that there is reasonable assurance that the systems affected by the proposed TS changes will perform their required safety functions.
3.1.3 In 10 CFR 50.49(b)(1 )(ii), design basis events are defined as conditions of normal operation, including AOOs, DBAs, external events, and natural phenomena for which the plant must be designed to ensure functions (b)( 1)(i) (A) through (C) of 10 CFR 50.49.
3.1.4 General Design Criterion (GDC) 10, "Reactor design," of Appendix A, "General Design Criteria for Nuclear Plants," to 10 CFR Part 50, requires in part that the reactor protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of ADOs.
3.1.5 GDC 13, "Instrumentation and control," of Appendix A to 10 CFR Part 50, requires that the instrumentation be provided to monitor variables and systems and that controls be provided to maintain these variables and systems within prescribed operating ranges during normal operation, ADOs, and accident conditions. Specifically, the staff reviewed the proposed TS changes and the affected instrument setpoint calculations and plant surveillance procedures to ensure proper operation of the HPCS/RCIC systems.
- 4 3.1.6 GDC 20, "Protection system functions," of Appendix A to 10 CFR Part 50, requires in part that the protection system be designed 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 AOOs. The staff evaluated the LAR to ensure that the proposed TS change will still protect the fuel design limits and plant SLs specified in GGNS TS 2.0 and that these SLs will not be exceeded under plant transient, AOOs, and accident conditions.
3.2 Regulatory Guidance:
3.2.1 Regulatory Guide (RG) 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," issued December 1999 (ADAMS Accession No. ML993560062),
describes a method acceptable to the NRC staff for complying with the agency's regulations for ensuring that setpoints for safety-related instrumentation are initially within and remain within the TS limits. GGNS is committed to using the guidance in RG 1.105. The RG endorses Part I of Instrument Society of America (ISA)-S67.04 1994, "Setpoints for Nuclear Safety Instrumentation," subject to the NRC staff clarifications. The staff used this guide to establish the adequacy of the GGNS setpoint calculation methodologies and the related plant surveillance procedures.
3.2.2 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 LSSSs assessed during testing and calibration of instrumentation. This RIS discusses why compliance to the AVs in the TSs during testing or calibration alone is not sufficient to ensure that the 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, it is recognized in RIS 2006-17 that other methods and approaches may also be acceptable.
The NRC staff used RIS 2006-17 to evaluate the effects of the proposed TS changes on the plant SLs, the acceptability of the setpoint calculation methodology, and the adequacy of the proposed TS changes to meet the requirements of 10 CFR 50.36.
RIS 2006-17 provides guidance for identifying functions on which SLs have been placed to meet the requirements of 10 CFR 50.36(c)(1 )(ii)(A). RIS 2006-17 specifically refers to Standard Technical Specifications Section 2.1.1, "Reactor Core SLs." GGNS TS 2.1.1.3 states, "Reactor vessel water level shall be greater than the top of active irradiated fuel,"
without mentioning any limitation on the TS mode or applicable condition.
RIS 2006-17 addresses the NRC staff position on LSSSs assessed during periodic testing and calibration of instrumentation. This RIS discusses issues that could occur during testing of LSSSs and which, therefore, may have an adverse effect on equipment operability.
- 5 3.2.3 Letter from Patrick L. Hiland, NRC, to NEI [Nuclear Energy Institute] Setpoint Methods Task Force, "Technical Specification for Addressing Issues Related to Setpoint Allowable Values," dated September 7,2005 (ADAMS Accession No. ML052500004).
The letter is complementary to RIS 2006-17.
4.0 TECHNICAL EVALUATION
The proposed changes correct nonconservative TSs, do not change the design or equipment associated with the transfer function of HPCS or RCIC from the CST to the suppression pool, and do not change the licensing basis with respect to the transfer of HPCS or RCIC from the CST to the suppression pool. Since the design of the transfer function does not change, the requirements of the above GDC continue to be met.
4.1 Proposed Changes to the TSs In its application, the licensee proposed changes to the AVs for the CST level-low functions for the operation of the HPCS system and RCIC system in TS Tables 3.3.5.1-1 and 3.3.5.2-1, respectively. These are the setpoints for the transfer of the HPCS and RCIC suction flow paths from the CST to the suppression pool. The licensee stated that the changes are necessary to correct an error in the original plant design in that the current TS values are non-conservative.
The licensee proposed to change the current AV (1) for the HPCS system in TS Table 3.3.5.1-1 from -3 inches to 4.7 feet, and (2) for the RCIC system in TS Table 3.3.5.2-1 from -3 inches to 3.7 feet. The licensee is not proposing to change any other requirement in these two TS tables.
There are no changes to the surveillance requirements (SRs), the LCOs, applicable modes, or the action conditions for the remedial actions if an LCO is not met.
In explaining the basis for the proposed changes, the licensee stated that the current CST level low signals for the HPCS and RCIC systems are initiated through two safety-related transmitters, with either transmitter being able to affect the automatic suction transfer. These transmitters are connected to the HPCS/RCIC suction line inside the Auxiliary Building and, thus, are not connected to the CST. It was discovered that these transmitters may not be capable of providing the CST level-low trip that would transfer the HPCS/RCIC suction from the CST to the suppression pool under all conditions because the transmitters have an uncorrected static head between the transmitters and top of the suction piping. This static head is normally offset by the inverted HPCS/RCIC suction nozzle inside the CST; however, this would not be the case if a seismic or other event were to occur that resulted in a failure of the non-safety related portion of the HPCS/RCIC suction piping. Revision of the CST low-level suction transfer AVs resolved this issue by raising the minimum water level in the CST above the top of the HPCS/RCIC suction piping. With this change, in the event of a failure of the HPCS/RCIC piping, the licensee stated that an automatic transfer from the CST to the suppression pool will occur.
4.2 Analysis of Licensee's Setpoint Calculations In its letter dated September 5, 2007, the licensee provided setpoint calculations applicable to the proposed TS changes. The NRC staff reviewed the setpoint calculation methodologies and found that the licensee calculated (1) the nominal trip setpoint (NTSP) by adding total loop
- 6 uncertainty (TLU) to the analytical limit (AL), and (2) the AV by adding loop uncertainty (LU) to the AL. To calculate TLU and LU, the licensee used the square root of the sum of the squares (SRSS) of the random variables and the summation of the independent variables. The licensee calculated TLU by adding SRSS of all drifts to LU.
The licensee used the new AL of 3 feet. The previous AL was based on the level of the vortex breaker of the suction piping. However, the portion of the suction piping that connects to the CST is non-safety related. To account for the suction pipe failure, the licensee used the top of the suction pipe as the AL in the revised setpoint calculations.
For CST level-low instrumentation for the HPCS system in TS Table 3.3.5.1-1, the licensee proposed to increase the AV from "~ - 3 inches" to "~ 4.7 ft." In calculation JC-Q1 E22-N654-1, the licensee calculated TLU to be 1.995 feet and LU to be 1.606 feet, based on which the licensee selected the NTSP to be 5 feet and the AV to be greater than or equal to 4.7 feet.
For CST level-low instrumentation for the RCIC system in TS Table 3.3.5.2-1, the licensee proposed to increase the AV from "~ - 3 inches" to "~ 3.7 ft." In calculation JC-Q1 E51-N635-1, the licensee calculated TLU to be 0.947 feet and LU to be 0.601 feet, based on which the licensee selected the NTSP to be 4 feet and the AV to be greater than or equal to 3.7 feet. In the letter dated September 5, 2007, the licensee stated that GGNS has specified 0.25 percent of the span as the as-found and as-left tolerances in the plant surveillance procedures for the affected instruments. Based on vendor data, the licensee used a reference accuracy of 0.25 percent of the span in calculating the TLUs and LUs. For a 40-feet span, this 0.25 percent equates to 0.1 foot. The NRC staff finds that the as-found and as-left tolerances are reasonably small (0.1 foot) compared to the TLU and LU tolerances, and the AVs are reasonably close to the NTSPs (NTSP - AV =0.3 foot).
The NRC staff compared the as-found and as-left tolerances to the guidance provided in the RIS 2006-17 and determined that the licensee's as-found and as-left values are conservative and consistent with the guidance in the RIS 2006-17. The staff also found that the GGNS setpoint calculation methodology is consistent with the guidance in RG 1.105. Furthermore, the proposed AVs will not only correct an error in the licensee's setpoint calculation, but are conservative because the AVs for the CST low level instrument channels are increased from
"~ - 3 inches" to "~4.7 ft." for HPCS and "~ - 3 inches" to "~3.7 ft." for RCIC.
The staff also reviewed the GGNS plant procedures for surveillance testing of the affected instrument channels and found that the licensee has applicable procedures to (1) check the as found values and take necessary steps to set the channels within acceptable tolerances and (2) take corrective actions if the procedural steps cannot be completed or the as-found values exceed the TS tolerance. The NRC staff found the licensee's procedures acceptable.
Based on the above evaluation, the NRC staff concludes that the proposed changes to the calculated AVs in the TS Tables 3.3.5.1-1 and 3.3.5.2-1 for the CST level-low setpoints for the automatic transfer to the suppression pool are acceptable. This is because the change in AVs will ensure that the suppression pool, which is the credited safety-related makeup water source, will be available to mitigate AOOs and postulated accidents.
- 7 4.3 Analysis of SL-Related Criteria Paragraph 50.36(c)(1 )(ii)(A) of 10 CFR Part 50 states that where an LSSS is specified for a variable on which a safety limit has been placed, the setting must be so chosen that the automatic protective action will correct the abnormal situation before an SL is exceeded. These LSSSs are designated SL-related LSSSs.
GGNS TS 2.0 specifies the SLs for the plant. The NRC staff evaluated the CST level-low setpoints for HPCS and RCIC to determine if they should be considered as SL-related. The SL of concern is TS 2.1.1.3, which requires that the reactor vessel water level shall be greater than the top of active irradiated fuel (TAF). UFSAR Section 6.3.1.1.1.c states that the HPCS system is capable of maintaining the water level above the top of the core.
The RCIC system is not an engineered safety feature system and no credit is taken by the licensee in any DBA for RCIC system operation. Hence, the automatic transfer function of the RCIC system from the CST to the suppression pool is not credited in any design basis event requiring safety-related SSCs.
However, the NRC staff concluded that the automatic transfer from the CST to the suppression pool is required for the HPCS system to comply with the plant design basis. Therefore, on this basis, the NRC staff concludes that the HPCS function (LSSS) for this automatic transfer is SL related. This is consistent with the statement in UFSAR Section 6.3.1.1.1.c that the HPCS system is provided for maintaining the water level above the top of the core. The staffs conclusion was further documented in a letter to the licensee from M. Markley (NRC) dated November 5,2008 (ADAMS Accession No. ML082490768). The licensee provided its response to the staffs conclusion by letter dated January 19, 2009 (ADAMS Accession No. ML090210237).
The licensee's amendment request does not resolve the generic issue raised by RIS 2006-17 with respect to LSSSs assessed during periodic testing and calibration of instrumentation. As discussed in RIS 2006-17, the NRC staff intends to resolve the issue with the industry through the Technical Specification Task Force (TSTF), and to incorporate setpoint issue guidance in the final approved TSTF traveler, which the staff then intends to issue as a consolidated line item improvement process. The NRC staff believes this will establish a uniform, satisfactory resolution that addresses the industry and staff concerns with instrument settings, conforms to the staffs operability guidance, and ensures compliance with 10 CFR 50.36. The licensee provided the following regulatory commitment in its letter dated February 20, 2009:
Entergy will follow the efforts of the Technical Specification Task Force (TSTF) and NRC to finalize the details and scope of the changes needed to resolve the instrument setpoint issue discussed in RIS-2006-17. If the Condensate Storage Tank (CST) Level-Low setpoint is affected by the approved TSTF traveler, then Entergy will submit a separate amendment request to implement the approved generic change within 6 months of approval of the TSTF Traveler.
In its review of the application, the NRC staff had several discussions with the licensee staff regarding whether or not the CST level setpoints are safety limit (SL)-related, in accordance with RIS 2006-17. The licensee's position is that the setpoints associated with the suction swap
- 8 for HPCS are not SL-related. The NRC staff disagrees, as documented in its letter to the licensee dated November 5, 2008. However, since (1) the setpoint methodology and surveillance testing procedures used by the licensee are consistent with the guidance of RIS 2006-17, and the proposed changes will (2) correct an error in the TSs by conservatively increasing the HPCS and RCIC suction swap values, (3) not present an immediate safety concern for current plant operation, and (4) end the prolonged administrative control of the TS values, the NRC staff finds the proposed changes acceptable. Resolution of the control of the CST swapover setpoints within TSs will be addressed pending final approval of the TSTF traveler, as documented in the licensee's letter dated February 20, 2009.
4.4 Conclusion Based on the above evaluation, the NRC staff concludes that the proposed changes to the AVs for the CST level-low setpoints for the automatic transfer to the suppression pool are acceptable. This is because the change in AVs will ensure that the suppression pool, which is the credited safety-related makeup water source, will be available to mitigate AOOs and postulated accidents. The proposed changes correct nonconservative TSs, do not change the design or equipment associated with the transfer function of HPCS or RCIC from the CST to the suppression pool, and do not change the licensing basis with respect to the transfer of HPCS or RCIC from the CST to the suppression pool. Therefore, the NRC staff further concludes that the proposed changes to the AVs in TS Tables 3.3.5.1-1 and 3.3.5.2-1 are acceptable.
The licensee's amendment request does not resolve the generic issue raised by RIS 2006-17 with respect to LSSSs assessed during periodic testing and calibration of instrumentation. The licensee's position is that the setpoints associated with the suction swap for HPCS are not SL related. The NRC staff disagrees, as documented in its letter to the licensee dated November 5, 2008. However, since (1) the setpoint methodology and surveillance testing procedures used by the licensee are consistent with the guidance of RIS 2006-17, and the proposed changes will (2) correct an error in the TSs by conservatively increasing the HPCS and RCIC suction swap values, (3) not present an immediate safety concern for current plant operation, and (4) end the prolonged administrative control of the TS values, the NRC staff finds the proposed changes acceptable. Resolution of the control of the CST swapover setpoints within TSs will be addressed pending final approval of the TSTF traveler, as documented in the licensee's letter dated February 20, 2009.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Mississippi 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 installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. 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
- 9 significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on May 8, 2007 (72 FR 26176). Accordingly, the amendments meet 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: T. Nakanishi S. Mazumdar C. Schulten Date:
February 25, 2009
- Previously concurred OFFICE NRR/LPL4/PM NRR/LPL4/LA DIRS/ITSB/BC DE/EICB/BC DSS/SRXB/BC OGC NRR/LPL4/BC NRR/LPL4/PM NAME FLyon JBurkhardt*
RElliott WKemper GCranston*
LSubin*
MMarkley FLyon DATE 2/24/09 1/30/09 2/25/09 2/24/09 2/23/09 2/18/09 2/25/09 2/25/09