ML20100H992
| ML20100H992 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/02/2020 |
| From: | Richard Guzman NRC/NRR/DORL/LPL1 |
| To: | Entergy Nuclear Operations |
| Guzman R | |
| References | |
| EPID L-2019-LLA-0262 | |
| Download: ML20100H992 (19) | |
Text
June 2, 2020 Vice President, Operations Entergy Nuclear Operations, Inc.
Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249
SUBJECT:
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 ISSUANCE OF AMENDMENT NO. 269 RE: TECHNICAL SPECIFICATION CHANGES TO CITY WATER SURVEILLANCE REQUIREMENT AND CONDENSATE STORAGE TANK REQUIRED ACTION A.1 (EPID L-2019-LLA-0262)
Dear Sir or Madam:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 269 to Renewed Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3 (Indian Point 3). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 21, 2019.
The amendment revises TS 3.7.7, City Water (CW), Surveillance Requirement 3.7.7.2, and TS 3.7.6, Condensate Storage Tank (CST), Required Action A.1, to allow one of the backflow preventer isolation valves on the Indian Point 3 city water header supply to be maintained closed when the steam generators are relied upon for heat removal, provided the requirements of TS Limiting Condition for Operation 3.7 are met.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286
Enclosures:
- 1. Amendment No. 269 to DPR-64
- 2. Safety Evaluation cc: Listserv
ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 269 License No. DPR-64
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Entergy Nuclear Operations, Inc. (ENO, the licensee) dated November 21, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.
Enclosure 1
- 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 Renewed Facility Operating License No. DPR-64 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 269, are hereby incorporated in the renewed license. ENO shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by James G.
James G. Danna Date: 2020.06.02 12:04:07 Danna -04'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: June 2, 2020
ATTACHMENT TO LICENSE AMENDMENT NO. 269 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 RENEWED FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Page Insert Page 3 3 Replace the following pages of 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.
Remove Pages Insert Pages 3.7.6 - 1 3.7.6 - 1 3.7.6 - 2 3.7.6 - 2 3.7.7 - 2 3.7.7 - 2
(4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Amdt. 203 to receive, possess, and use in amounts as required any 11/27/00 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components.
(5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Amdt. 203 possess, but not separate, such byproduct and special 11/27/00 nuclear materials as may be produced by the operation of the facility.
C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 below:
(1) Maximum Power Level ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 3216 megawatts thermal (100% of rated power).
(2) Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 269, are hereby incorporated in the Renewed License. ENO shall operate the facility in accordance with the Technical Specifications.
(3) (DELETED) Amdt. 205 2-27-01 (4) (DELETED) Amdt. 205 2-27-01 D. (DELETED) Amdt. 46 2-16-83 E. (DELETED) Amdt. 37 5-14-81 F. This amended license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter of May 2, 1975, to Consolidated Edison Company of New York, Inc., granting a Section 401 certification under the Federal Water Pollution Control Act Amendments of 1972.
Amendment No. 269
CST 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Condensate Storage Tank (CST)
LCO 3.7.6 The CST shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CST inoperable. ---------------------NOTE---------------------
OPERABILITY of City Water with CST inoperable requires Unit 3 City Water Header Supply Valves to be open.
A.1.1 Open closed Unit 3 City Immediately Water Header Supply Isolation Valve if previously closed per Note modifying SR 3.7.7.2.
AND A.1.2 Verify by administrative Immediately means OPERABILITY of City Water. AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND A.2 Restore CST to OPERABLE. 7 days (continued)
INDIAN POINT 3 3.7.6-1 Amendment 269
CST 3.7.6 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4, without 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> reliance on steam generator for heat removal.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the CST level is 360,000 gal. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> INDIAN POINT 3 3.7.6-2 Amendment 269
CITY WATER 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify the CW header pressure is 30 psig. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.7.7.2 ------------------------------------NOTE----------------------------------
One backflow preventer isolation valve on the Unit 3 City Water Header Supply may be maintained closed, provided the requirements of LCO 3.7.6 are met.
For Unit 3 City Water Header Supply Isolation valves 31 days maintained open, verify valves are open.
AND For one backflow preventer isolation valve on the Unit 3 31 days City Water Header Supply maintained closed, verify capability of valve to be opened.
SR 3.7.7.3 Perform testing required by Inservice Testing Program for In accordance with each valve needed to align CW to each AFW pump suction. the Inservice Testing Program INDIAN POINT 3 3.7.7-2 Amendment 269
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 269 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-64 ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
1.0 INTRODUCTION
By letter dated November 21, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19325E913), Entergy Nuclear Operations Inc. (the licensee) submitted a license amendment request (LAR or the application) to revise the Technical Specifications (TSs) for Indian Point Nuclear Generating Unit No. 3 (Indian Point 3 or IP3).
The amendment would revise TS 3.7.6, Condensate Storage Tank (CST), and TS 3.7.7, City Water (CW), surveillance requirements (SRs). Specifically, the proposed change to Indian Point 3 TS 3.7.6 Required Action A.1 relates to additional actions that would be required if the CST is inoperable, and the proposed change to SR 3.7.7.2 would allow one of the backflow preventer isolation valves on the CW header supply to be maintained closed when in the modes of applicability for TS Limiting Condition for Operation (LCO) 3.7.6.
2.0 REGULATORY EVALUATION
2.1 System Description The Indian Point 3 auxiliary feedwater (AFW) system automatically supplies feedwater to the steam generators (SGs) to remove decay heat from the reactor coolant system (RCS) upon the loss of normal feedwater supply. The AFW pumps take suction from the CST and CW.
The normal (preferred) water supply for the AFW system is the CST, which provides a safety-grade source of water to the SGs for removing decay and sensible heat from the RCS. A backup cooling water supply to the AFW pumps is provided by the CW system. When the CST is unavailable, or when the CST supply is exhausted, CW is used to supply the AFW system for decay heat removal and plant cooldown.
Enclosure 2
Because the CST is a principal component in removing residual heat from the RCS via the AFW system, it is a seismic Category I structure designed to withstand earthquakes and other natural phenomena. However, the CST is not designed to withstand the effects of a tornado-generated missile. As a result, the AFW system is provided with redundant water supplies such that an alternate source of water from the CW tank is available in the event the CST is damaged by a tornado-generated missile. Only when the CST supply is exhausted or not available will CW be used to supply the AFW system.
2.2 Description of the Proposed Changes This amendment would modify TS 3.7.6 and TS 3.7.7 SRs. The proposed change to SR 3.7.7.2 would allow one of the existing backflow preventer isolation valves on the CW header supply to be maintained closed when in the modes of applicability for TS LCO 3.7.7.
As indicated in the LAR, the basis for the proposed amendment is to stop CW intrusion into the AFW system and CST due to leak-by past a downstream isolation valve. This leak-by has created water chemistry concerns, particularly for the SGs, for which the CST is the preferred water source. The proposed changes would also allow removal of a temporary modification that provides continuous flushing of the AFW pump suction line associated with the valve leak-by to alleviate water chemistry concerns.
2.2.1 Proposed Change to TS LCO 3.7.6 The operability of the CST is determined by maintaining the tank volume at or above the minimum required volume. LCO 3.7.6 requires CST volume greater than 360,000 gallons. The LCO is applicable in Modes 1, 2, and 3, and in Mode 4 when the SGs are relied upon for heat removal.
The licensee proposed changes to TS 3.7.6, Required Action A.1, to require a closed backflow preventer isolation valve on the CW header supply to be reopened immediately in the event the CST is declared inoperable. In addition, a note to Required Action A.1 would be added to specify that operability of CW with the CST inoperable requires the Unit No. 3 CW header supply valves to be open.
2.2.2 Proposed Change to TS 3.7.7.2 SR for CW Three valves are subject to SR 3.7.7.2. Under the current requirements of TS 3.7.7, all three valves are maintained open when in the modes of applicability for LCO 3.7.7 (i.e., during Modes 1, 2, and 3, and Mode 4 when the SGs are relied upon for heat removal) and are required to be verified open once every 31 days in accordance with SR 3.7.7.2 requirements.
Due to the continuing concerns related to CW intrusion into the AFW system and the CST, this request proposes a change to SR 3.7.7.2 to allow one of the backflow preventer isolation valves on the Indian Point 3 CW header supply to be maintained closed when in the modes of applicability for LCO 3.7.7 (i.e., during Modes 1, 2, and 3, and Mode 4 when the SGs are relied upon for heat removal), provided the requirements of LCO 3.7.6 are met.
2.3 Regulatory Requirements and Guidance Used in the Evaluation of the Changes Title 10 of the Code of Federal Regulations (10 CFR) establishes the fundamental regulatory requirements. In 10 CFR 50.36, the U.S. Nuclear Regulatory Commission (NRC or the
Commission) established its regulatory requirements related to the content of the TSs. The regulatory requirements upon which the NRC staff based its review of the application are as follows:
The regulation at 10 CFR 50.36(c)(2) states that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
Criterion 3 of 10 CFR 50.36(c)(2) requires that a TS LCO 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 regulation at 10 CFR 50.36(c)(3) requires that SRs shall be included in the TSs and shall include 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.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 18, Revision 3, Human Factors Engineering, provides the NRC staffs guidance for the review of human performance for applicants, and Chapter 13.5.2.1, Revision 2, Operating and Emergency Operating Procedures, provides the methodology for the NRC staffs review of operating procedures that will be used by the operating organization to ensure that routine operating, off-normal, and emergency activities are conducted in a safe manner.
NUREG-0711, Human Factors Engineering Program Review Model, Revision 3, provides the methodology for the NRC staffs review of human factors engineering programs.
3.0 TECHNICAL EVALUATION
3.1 Changes to TS LCO 3.7.6, Required Action A.1 and SR 3.7.7.2 The NRC staff evaluated the possible impact on availability of inventory for AFW safety function, the accessibility of the closed valves, and the likelihood of CW being required. The NRC staff further assessed the potential of a CST failure requiring AFW as the alternate source of inventory.
Indian Point 3 is equipped with an AFW system, which supplies feedwater to the secondary side of the SGs when the normal feedwater supply is not available. The system removes decay heat from the reactor core by heat exchange in the SG when the main feedwater pumps are not functional, thereby maintaining the required heat sink for the RCS. The AFW system also functions as an engineered safeguards system and is directly relied upon to remove reactor decay heat from the RCS and prevent core damage and system-overpressurization in the event of transients and accidents. The normal source of inventory for the AFW system is the CST (LCO 3.7.6). In the event that the CST is unavailable, CW inventory is used as a backup water supply to the AFW pumps.
The AFW system consists of two motor-driven pumps and one turbine-driven pump and associated valves, piping, and controls to enable the AFW system to satisfy single failure and diversity considerations. In addition to the redundancy afforded by the AFW pumps, the water supply source for the AFW system is also redundant. Each pump is normally aligned to CST as
the primary source. In accordance with the plant design basis, the CST is the safety-related water source for the AFW system and contains enough usable inventory to enable the AFW system to cool the plant to hot standby conditions and maintain the plant in hot standby for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a reactor trip from full power operation, concurrent with a loss-of-offsite-power transient.
The CST provides a passive flow of water, by gravity, to the AFW system. It has a capacity of 600,000 gallons and is sized to meet the plants normal operating and maintenance needs with 360,000 gallons of CST inventory dedicated to support AFW operation. To ensure that adequate inventory is maintained in the CST to support the AFW safety function, TS 3.7.6 requires a minimum inventory of 360,000 gallons to be maintained in the CST. When the CST supply is exhausted, CW will be used. Operability of the CST is determined by maintaining the tank volume at or above the minimum required volume.
CW is the backup or alternate source to the CST as a water supply for the AFW system. The CW system includes the site CW header consisting of the 1.5 million gallon CW tank and connection to the offsite water supply. Operability of the CW system is determined by maintaining the tank volume at or above the minimum required volume and periodic verification that the CW header supply isolation valves are open.
During normal operation, CW is isolated from the AFW pump suction by use of normal isolation valves in the AFW pump supply piping. Currently, CW is placed in service by opening these individual AFW pump isolation valves from the control room, (i.e., CT-PCV-1187, CT-PCV-1188, and CT-PCV-1189), which represent the boundary between CW and the AFW pump suction. In addition to these isolation valves, CW lines contain normally open backflow preventer isolation valves in the supply lines to avoid backflow from the AFW system into CW. Current TSs require all isolation valves of the CW backflow preventer to be maintained open when in the modes of applicability for LCO 3.7.7 and are required to be verified open once every 31 days in accordance with SR 3.7.7.2. These backflow preventer isolation valves include CT-49 (Unit No. 3 CW supply header isolation valve), CT-1300 (CW inlet to CT-1301 backflow preventer isolation), and CT-1302 (CW outlet to CT-1301 backflow preventer isolation).
The requested change adds an additional manual action to align the alternate CW supply for the AFW pumps when the normal source of CST water is unavailable or when the tank level falls below the minimum required level. The LAR proposes changes to LCO 3.7.6 and SR 3.7.7.2 to allow one of the backflow preventer isolation valves on the CW header supply to be maintained closed in the modes of applicability for LCO 3.7.7 (i.e., during Modes 1, 2, and 3, and Mode 4 when the SGs are relied upon for heat removal), provided the requirements of LCO 3.7.6 are met. This revised TS will also require a closed backflow preventer isolation valve on the CW header supply to be reopened immediately in the event the CST is declared inoperable. This realignment of CW to the suction of the AFW pumps will require reopening the manually closed backflow preventer isolation valve by a nuclear plant operator, in addition to the already required manual action of opening the individual AFW pump suction remote air-operated valves from the control room. Therefore, this request does not replace any automatic action with a manual action, isolate a safety-related source of water that was previously unisolated, or close a valve that was previously required to be locked open.
Both CW backflow preventer isolation valves CT-1302 and CT-1301 are located in an area accessible within the AFW room. The AFW pump room and the equipment located in this room are protected from the effects of tornadoes and tornado-generated missiles. The backflow preventer isolation valves on the CW header supply are 6-inch manually operated gate valves
and are not in a harsh or corrosive environment that would degrade or affect their condition.
Therefore, operator access to these valves is not expected to be impeded when needed, since an acceptable ambient environment is normally present in the AFW pump room.
The CST is a principal component in removing residual heat from the RCS. The CST is designed to withstand earthquakes and other natural phenomena but is not designed to withstand the effects of a tornado-generated missile. Therefore, the greatest risk of failure that would require CW would be failure of CST from a tornado missile strike. The licensee is crediting the CW as a redundant source for addressing this failure scenario of the CST. Since CW provides the AFW with a backup water source should the CST fail, supply water to the AFW pumps will remain available. Therefore, the risk associated with this change is primarily due to the potential unavailability of CST following tornado events. The probability of a tornado strike on this plant is generally considered low, and the probability of missile impact failure of the CST would be lower. Therefore, the risk of failure of CST due to tornado missile and subsequent inability to provide CW inventory is determined to be very small. As a result, this change has a very small impact on safety margin and defense in depth.
The proposed amendment will not involve any physical changes to the existing plant, and the plant retains its current piping configuration to use CW as the AFW alternate water source. The primary source of CST cooling water to the AFW pumps is also unaffected by the proposed changes. The proposed amendment makes no change to conditions external to the plant that could create the possibility of a new or different kind of accident and does not impact the availability of CW inventory. Further, the reconfiguration of the valve position to allow manually closed isolation will have little impact on AFW normal operating safety function. As a result, the AFW system, the CST, and the CW system will continue to perform their support functions for previously evaluated accidents or events for which the AFW system is credited or required.
The licensee indicated it will revise Standard Operating Procedure 0-SOP-WEATHER-002, Severe Weather Preparations, to require a nuclear plant operator to reopen the backflow preventer isolation valve on the Indian Point 3 CW header supply in the event a tornado warning is issued by the National Weather Service as a result of severe weather conditions. To further minimize any increase in risk, the licensee stated that procedure controls will require the reopening (or verification of open status) of the closed backflow preventer isolation valve as part of the actions required in response to a CST trouble alarm, CST lo-lo level alarm, issuance of a tornado warning from the National Weather Service, inoperable CST, or as part of placing the CW system in service for conditions in which the control room is unavailable or in response to a loss of secondary heat sink.
Criterion 3 of 10 CFR 50.36(c)(2) requires that a TS LCO 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. Based on the above evaluation, the NRC staff concludes that revision of LCO 3.7.6 Required Action A.1 meets the requirements of 10 CFR 50.36 for inclusion in the plants TSs. The proposed change to LCO 3.7.6 continues to provide assurance that the lowest functional capability or performance levels of equipment required for safe operation of the facility will be met. Therefore, the LCOs will continue to meet 10 CFR 50.36(c)(2).
In addition to the proposed revision of LCO 3.7.6, the LAR also proposes corresponding changes to TS SR 3.7.7.2 to allow one of the backflow preventer isolation valves on the Indian Point 3 CW header supply to be maintained closed when in the modes of applicability for TS
LCO 3.7.7 (i.e., during Modes 1, 2, and 3, and Mode 4 when the SGs are relied upon for heat removal), provided the requirements of LCO 3.7.6 are met.
As stated in 10 CFR 50.36(c)(3), the SRs in a TS are to demonstrate that the LCO in that specification is being met. The regulation at 10 CFR 50.36(c)(3) identifies that SRs shall be included in the TSs and shall include 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.
The proposed changes to SR 3.7.7.2 and LCO 3.7.6, Required Action A.1, would continue to require the valves that are maintained open to be verified that they are open every 31 days. In addition, the SR would require verification that a backflow preventer isolation valve maintained closed is capable of being opened under the same 31-day frequency. This additional surveillance will provide assurance that these valves are able to perform their function.
Based on the above evaluation, the NRC staff concludes that the addition of SR 3.7.7.2 demonstrates that LCO 3.7.7 is being met and, therefore, meets the requirements of 10 CFR 50.36 for inclusion in the plants TSs.
3.2 Manual Operator Actions The licensees proposed change to TS 3.7.6, Required Action A.1, would require the closed backflow preventer isolation valve on the Indian Point 3 CW header supply to be reopened immediately in the event the CST is declared inoperable. This is a new manual action that would locally dispatch an operator to the AFW pump room.
The licensee states that the AFW pump room is easily accessible by the operator with no impediments. The backflow preventer isolation valves located in the AFW pump room are 6-inch manually operated gate valves and are easily accessible by the operator. The environment inside the pump room will not degrade or affect the condition of the valves.
Additionally, the AFW pump room and equipment located in this room are protected from the effects of tornadoes and tornado-generated missiles unlike the CST. While the CST is the preferred source of water for the SGs, it is designed to withstand the effects of only a design-basis earthquake. Subsequently, in the event a tornado-generated missile damages the CST, CW is the non-safety-related backup to the CST as a water supply for the AFW system.
The likelihood of a high energy line break or fire in the AFW pump room concurrent with a CST malfunction or a tornado-generated missile is highly unlikely and not in the current licensing basis for Indian Point 3. In the event immediate access to the AFW pump room is impacted, accessibility would be quickly restored.
The licensee provided feedback from Indian Point 3 operations regarding operator action timing.
As stated in Attachment 4, Risk Assessment for Proposed Changes to IP3 Technical Specifications (TS), Section 3.2.2, Tornado Risk Due to Missiles, of the licensees application, the time required to align the CW backup to AFW is estimated to be an additional 15 minutes.
Furthermore, by allowing one of the backflow preventer isolation valves to be closed during modes of applicability for TS LCO 3.7.7, the CW intrusion into the AFW system and CST will be substantially reduced. The proposed change to Indian Point 3 TS SR 3.7.7.2, in addition to TS 3.7.6, Required Action A.1, will allow the temporary modification of the continuous flushing of the 33 AFW pump suction piping to be removed.
Based on the considerations above, the NRC staff finds the licensee has provided sufficient information for the new operator action and finds it to be acceptable.
3.3 Task Analysis In Section 2 of the enclosure to the LAR, the licensee identified several level alarms for the CST available to Indian Point 3 operators. The licensee does not propose adjusting the setpoints to actuate the alarms, but does propose a manual operator action in response to these alarms. As stated in Section 3.2 above, the proposed change to TS 3.7.6, Required Action A.1, would require the closed backflow preventer isolation valve to be reopened immediately in the event the CST is declared inoperable. Specifically, this is a new manual operator action that aligns the backup non-safety-related CW supply for the AFW pumps when the normal source of CST water is unavailable or depleted.
The first alarm identified by the licensee is the CST trouble alarm and is associated with Alarm Response Procedure 3-ARP-006, Panel SCF-Condensate and Feedwater. This alarm is a TS monitoring alarm that alerts the operators of depletion in the normal CST inventory and is set to 20.66 feet. This setpoint is above the minimum TS volume of water (360,000 gallons; maximum is 600,000 gallons) required by TS SR 3.7.6.1. The licensee states that there are approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> available to reopen the valve once the CST trouble alarm is received prior to reaching the lo-lo level alarm, providing the operator with sufficient time to open the valve.
The second alarm identified is the two redundant low-low level alarms that are set at 2.75 feet and are associated with Alarm Response Procedure 3-ARP-013, Panel SKF-Bearing Monitor.
These alarms are intended to alert the operators that the preferred source of water for the suction of the AFW pumps has been depleted and require actions to place the CW system in service. The licensee states that a closed valve is unlikely, and there is sufficient time available for the operator to reopen the valve and place the CW system in service prior to reaching the CST levels where air ingestion into the AFW pump suction would be of concern.
In addition to these alarms, the licensee states that the design of the CST provides for automatic isolation of the condensate system makeup supply by closing valves if the CST level decreases below 19.66 feet, to ensure the minimum TS volume of water.
Based on the considerations above, the NRC staff finds that the licensee has appropriately addressed the task analysis for the proposed license amendment and finds it to be acceptable.
3.4 Procedure Design In Section 2 of the enclosure to the LAR, the licensee identified several procedures that will be revised as a result of the proposed amendment.
The first set of revisions is for plant procedures to assure that sufficient time is available to align the CW system as a backup source of water for the AFW pumps in the event the CST is unavailable or depleted. As discussed in Section 3.3 above, these procedures are: (1) Alarm Response Procedure 3-ARP-006, Panel SCF-Condensate and Feedwater, associated with the CST trouble alarm, and (2) 3-ARP-013, Panel SKF-Bearing Monitor, associated with the CST lo-lo level alarms.
Alarm Response Procedure 3-ARP-006 will require that the backflow preventer isolation valve on the Indian Point 3 CW header supply (i.e., either CT-1300 or CT-1302) be reopened when
this alarm is actuated in the control room. Alarm Response Procedure 3-ARP-013 will verify the closed backflow preventor isolation valve (i.e., either CT-1300 or CT-1302) has been opened as part of the actions to place the CW system in service.
The second set of revisions the licensee identified is for Standard Operating Procedures and an Emergency Operating Procedure. Standard Operating Procedure 0-SOP-WEATHER-002, Severe Weather Preparations, will be revised to require an operator to reopen the backflow preventer isolation valve on the Indian Point 3 CW header supply in the event a tornado warning is issued by the National Weather Service. The licensee identifies this change as a preemptive action because the CST is not protected from the effects of tornado-generated missiles, unlike the AFW pump room.
The licensee will also revise 3-SOP-ESP-01, Local Equipment Operation and Contingency Actions as part of the implementation of the proposed amendment. The revision to this procedure is necessary to require the closed backflow preventer isolation valves to be opened, in addition to the individual AFW pump suction valves, as part of the actions needed to place the CW system in service as the suction water supply to the AFW pump for conditions in which the control room is unavailable. EOP 3-FR-H.1, Response to Loss of Secondary Heat Sink will be revised to respond to a loss of secondary heat sink under the Function Restoration Procedures for the Emergency Operating Procedures.
Further, the licensee states that the affected sections of the Updated Final Safety Analysis Report will be revised as necessary.
Based on the considerations above, the NRC staff finds that the licensee has appropriately identified and addressed the procedure design for the proposed license amendment and finds it to be acceptable.
3.5 Human Reliability Analyses - Risk Important Human Actions In Attachment 4, Risk Assessment for Proposed Changes to IP3 Technical Specifications (TS), of the enclosure to the LAR, the licensee provides a probabilistic risk assessment evaluation to assess the impact on the Indian Point 3 risk of the proposed changes to TS SR 3.7.7.2 and TS 3.7.6, Required Action A.1. However, the human error probability associated with the probabilistic risk assessment will be not discussed in this evaluation based on the low safety significance determination as discussed in Section 3.6 below.
3.6 Risk Insights The licensee provided the basis for the proposed changes in Section 2 of the Enclosure to the LAR to establish the acceptability of the proposed changes to SR 3.7.7.2 and TS 3.7.6, Required Action A.1. In addition to the basis provided in Section 2, the licensee performed a risk assessment in support of the proposed changes and provided the associated assumptions, risk evaluation, and results in Attachment 4 of enclosure to the LAR.
The licensees risk evaluation was not provided consistent with the principles of risk-informed integrated decision-making in Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.
Because this is not a risk-informed application, the probabilistic risk assessment models, assumptions, and analyses used to derive risk insights in Attachment 4 of the enclosure to the LAR were not reviewed by the NRC staff to determine their technical acceptability as a basis to
support this application. As a result, the NRC staff did not rely on the numerical results provided by the licensee. However, the NRC staff considered the licensee-provided risk insights to aid in the deterministic review of the proposed changes. The NRC staff also performed an independent assessment using Indian Points Standardized Plant Analysis Risk (SPAR) Models to evaluate the risk contribution. The generic risk insights considered by the NRC staff supported the engineering conclusions that the proposed TS changes are of low safety significance.
3.7 NRC Staff Conclusion
The NRC staff has concluded, based on the considerations discussed in Section 3.0 above, that there is reasonable assurance that the requirements of 10 CFR 50.36 will continue to be met since the CW ability to provide redundant supply will not be affected by the proposed changes.
The NRC staff also found that there is reasonable assurance, given the ability to open subject isolation valves in an accessible location, verification of valve ability to open and close, procedural controls, and the fact that allowing one manual isolation valve in the closed state does not result in more than minimal increase in risk, that such activities will be conducted in compliance with the Commissions regulations. Therefore, the NRC staff finds the proposed changes acceptable.
4.0 REGULATORY COMMITMENT The licensee submitted the following Regulatory Commitment No. NL-19-093-01 in of the enclosure to the LAR. Regulatory commitments specify the items which the licensee volunteers to perform in support of its licensing application. Regulatory commitments do not require prior NRC approval of subsequent changes and, therefore, they are not enforceable licensing requirements. In its review of license applications, the NRC staff does not use the regulatory commitments as a basis in the safety evaluation for approving license amendments.
Type (check one) Scheduled Commitment One-Time Continuing Completion Action Compliance Date NL-19-093-01:
Revise plant procedures to require the re-opening (or verification of open status) of the closed backflow preventer isolation valve (CT-1300 or CT-1302) as part of actions As part of required in response to a CST Trouble Alarm, CST Lo-Lo Level approved LAR Implementation Alarm, issuance of a tornado warning from the National Weather Service, inoperable CST, or as part of placing CW System in service for conditions in which control room is unavailable or in response to a loss of secondary heat sink.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on February 28, 2020. 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 and changes SRs.
The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (January 17, 2020, 85 FR 3081). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: G. Curran D. Ki M. Reisi Fard Z. Coffman Date: June 2, 2020
ML20100H992 *by e-mail OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/SCPB/BC*
NAME RGuzman LRonewicz BWittick DATE 5/13/2020 5/13/2020 3/20/2020 OFFICE NRR/DRO/IOLB/BC* NRR/DRA/APLC/BC* NRR/DSS/STSB/BC*
NAME CCowdrey SRosenberg VCusumano DATE 5/11/2020 3/06/2020 5/15/2020 OFFICE OGC - NLO* NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME KGamin JDanna RGuzman DATE 5/26/2020 6/01/2020 6/02/2020