ML13326A047

Issuance of Amendment Connection of Non-Seismic Boric Acid Recovery System to the Refueling Water Storage Tank

ML13326A047
Person / Time
Site:	Indian Point
Issue date:	12/20/2013
From:	Pickett D Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Pickett D
References
TAC MF1440, FOIA/PA-2016-0148
Download: ML13326A047 (21)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 20, 2013 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.2- ISSUANCE OF AMENDMENT RE: CONNECTION OF NON-SEISMIC BORIC ACID RECOVERY SYSTEM TO THE REFUELING WATER STORAGE TANK (TAC NO. MF1440)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 273 to Facility Operating License No. DPR-26 for the Indian Point Nuclear Generating Unit No. 2. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated April15, 2013, as supplemented by letter dated September 4, 2013.

The amendment revises TS 3.5.4, "Refueling Water Storage Tank (RWST)," to allow for the temporary connection between the non-seismically qualified piping of the Boric Acid Recovery System to the seismically qualified piping of the RWST for the purpose of purifying the contents of the RWST in advance of the spring 2014 refueling outage. Operation in this mode will be under administrative controls and will only be applicable for limited periods through the end of the spring 2016 refueling outage.

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

Sincerely, Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-247

Enclosures:

1. Amendment No. 273 to DPR-26

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 2. LLC ENTERGY NUCLEAR OPERATIONS. INC.

DOCKET NO. 50-247 INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS Amendment No. 273 License No. DPR-26

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

A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated April 15, 2013, as supplemented on September 4, 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 set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-26 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendices A, Band C, as revised through Amendment No. 273, are hereby incorporated in the 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 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: December 20, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 273 FACILITY OPERATING LICENSE NO. DPR-26 DOCKET NO. 50-247 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 3 3 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Pages Insert Pages 3.5.4-1 3.5.4-1

instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Arndt. 42 to receive, possess, and use in amounts as required any 10-17-78 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 Arndt. 220 possess, but not separate, such byproduct and special 09-06-01 nuclear materials as may be produced by the operation of the facility and Indian Point Nuclear Generating Unit No. 3 (IP3).

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 1: 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; 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 ENO is authorized to operate the facility at steady state Arndt. 241 reactor core power levels not in excess of 3216 10-27-2004 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A, 8, and C, as revised through Amendment No. 273, are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications.

(3) The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:

1. This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee-controlled documents as described in Table R, "Relocated Technical Specifications from the CTS," and Table LA, "Removed Details and Less Restrictive Administrative Changes to the CTS" attached to the NRC staffs Safety Evaluation enclosed with this amendment. The relocation of requirements and detailed information shall be completed on or before the implementation of this amendment.

Amendment No. 273

RWST 3.5.4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

-NOTE-The RWST isolation valves 350, 727A and 845 connected to non-safety related piping may be opened under administrative controls for up to 30 days per fuel cycle for filtration until the end of refuel outage 22.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RWST boron A.1 Restore RWST to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> concentration not within OPERABLE status.

limits.

OR RWST borated water temperature not within limits.

B. One of the two required B.1 Restore RWST level low 7 days channels of the RWST low alarm to OPERABLE level low low alarm status.

inoperable.

C. RWST inoperable for C.1 Restore RWST to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reasons other than OPERABLE status.

Condition A or B.

D. Required Action and D.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 0.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> INDIAN POINT 2 3.5.4- 1 Amendment No. 273

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 273 TO FACILITY OPERATING LICENSE NO. DPR-26 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO.2 DOCKET NO. 50-247

1.0 INTRODUCTION

By letter dated April 15, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13116A007), as supplemented by letter dated September 4, 2013 (ADAMS Accession No. ML13253A138), Entergy Nuclear Operations, Incorporated requested a license amendment for the Indian Point Nuclear Generating Unit No. 2 (IP2) Technical Specifications (TSs). The proposed change would revise TS 3.5.4, "Refueling Water Storage Tank {RWST)," so that the non-seismically qualified piping of the temporary Boric Acid Recovery System (BARS) may be connected to the RWST seismic piping. Operation of the BARS from the RWST will be under administrative controls for a limited period of time {i.e.,

30 days for RWST filtration prior to each fuel cycle). This change is only applicable until Refueling Outage R22 ends (spring 2016).

Historically, IP2 was periodically connecting the non-seismic reverse osmosis system, known as the BARS, to the seismic spent fuel pool (SFP) purification loop to filter the RWST water while in plant conditions and modes for which the RWST was required to be operable. The licensee used this alignment prior to refueling outages to remove silica and filter the RWST water.

During refueling outages, RWST water is used to flood the refueling cavity to support fuel movement. Filtering the RWST water in advance of the refueling outage is highly desirable in order to remove suspended solids and improve water clarity so that refueling bridge operators can clearly identify fuel assemblies prior to fuel movement. Filtering the RWST contents can also reduce the activity and reduce accumulated dose to plant personnel. Prior to refueling outage (RFO) 2R20, the RWST was recirculated to the reverse osmosis system for a total of 13 days resulting in a silica concentration of less than 1.1 parts per million (ppm). Similarly, prior to RFO 2R 19, the RWST was recirculated for a total of 11 days resulting in a silica concentration of 1.3 ppm.

This operation was discontinued following the issuance of Nuclear Regulatory Commission (NRC) Information Notice (IN) 2012-01 "Seismic Considerations- Principally Issues Involving Tanks," on January 26, 2012 (ML11292A175). The IN provided examples and references to events in which licensees failed to recognize various seismic considerations and system alignment issues that could impact safety. Examples were identified in which licensees failed to recognize that aligning non-seismic piping to the RWST would require TS Limiting Conditions for Operation action statement entry, system modifications, or license amendments. The IN noted that the TSs would not allow applying compensatory measures, such as manual operator

actions in place of the closed boundary valve, for periods longer than the TS completion time for restoring the RWST to operable status, unless the TSs expressly permit operation under such measures.

The licensee's letter of April 15, 2013, recognizes the circumstances described in the IN, including the need to apply for an amendment of TS 3.5.4 so that the TS would expressly permit operation under the specific circumstances of RWST recirculation for filtration to improve the quality of the water for refueling activities. Accordingly, the licensee submitted this license amendment request to seek NRC approval to permit non-seismically qualified piping of the BARS to be connected to the Refueling Water Storage Tank seismic piping under administrative controls for a limited period of time in order to purify the contents of the RWST. This would be accomplished by realigning manual valves that are designated as American Society of Mechanical Engineers (ASME) code boundary valves. The piping configuration would only be permitted for a maximum of 30 days prior to the spring 2014 and 2016 refueling outages for IP2.

Following the spring 2016 refueling outage, a permanent modification will be completed so that connection of seismic and non-seismic piping would not be necessary to filter the RWST water.

The licensee has proposed to add the following note toTS 3.5.4:

The RWST isolation valves 350, 727 A and 845 connected to non-safety related piping may be opened under administrative controls for up to 30 days per fuel cycle for filtration until the end of refuel outage 22.

The supplemental letter dated September 4, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration.

2.0 REGULATORY EVALUATION

The following explains the applicability of General Design Criteria (GDC) for IP2. The construction permit for IP2 was issued by the Atomic Energy Commission (AEC) on October 14, 1966, and the operating license was issued on September 28, 1973. The plant GDC are discussed in the Updated Final Safety Analysis Report (UFSAR) Chapter 1.3, "General Design Criteria," with more details given in the applicable UFSAR sections. The AEC published the final rule that added Title 10 of the Code of Federal Regulations ( 10 CFR) Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223- Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the Appendix A GDC to plants with construction permits issued prior to May 21, 1971. Therefore, the GDC which constitute the licensing bases for IP2 are those in the UFSAR.

As discussed in the UFSAR, the licensee for IP2 has made some changes to the facility over the life of the unit that has committed to some of the GDCs from 10 CFR Part 50, Appendix A The extent to which the Appendix A GDC have been invoked can be found in specific sections of the UFSAR and in other IP2 licensing basis documentation, such as license amendments.

The regulatory requirements and guidance which the NRC staff considered in its review of the license amendment request (LAR) are as follows:

Title 10 of the Code of Federal Regulations (1 0 CFR) 50.67, "Accident source term" states that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE), (ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE), and (iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

Appendix A to 10 CFR Part 50, "General Design Criteria (GDC)," Criterion 19--Control room, states:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Equipment at appropriate locations outside the control room shall be provided ( 1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

GDC 2, "Design bases for protection against natural phenomena," requires that structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without the loss of the capability to perform their safety functions.

GDC 35, "Emergency core cooling," requires, in part, that a system provide abundant emergency core cooling. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Information Notice 2012-01 communicated operating experience involving operability of the RWST at the Shearon Harris Nuclear Power Plant. The seismically qualified RWST was aligned to the non-seismically qualified fuel pool purification system for purification of the RWST contents, creating a breach of the seismically qualified boundary and resulting in the inoperability of the RWST. The plant credited operator action, if needed to close the open valve at the seismically qualified boundary and declared the RWST operable for the duration of the

planned purification activity. It was determined by the NRC staff that, while entry into a TS action statement is allowable for maintenance or surveillances, the TS does not allow compensatory measures to be credited for periods longer than the TS completion time unless the TS expressly allows operation in that condition.

Information Notice 91-56, "Potential Radioactive Leakage to Tank Vented to Atmosphere,"

dated September 19, 1991 communicated that the engineered safety features (ESF) system could potentially leak through valves that isolate interfacing systems with the ESF systems to tanks that vent to the atmosphere. These tanks include the RWST.

Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Rev. 0, July 2000 provides the methodology for analyzing the radiological consequences of several design basis accidents to show compliance with 10 CFR 50.67. The RG 1.183 provides guidance to licensees on acceptable application of alternate source term (AST) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

The NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition." Chapter 13 addresses "Conduct of Operation," specific sub-chapters considered in this review were Chapters 13.2.1, "Reactor Operator Requalification Program; Reactor Operator Training," Rev. 3; and 13.5.2.1, "Operating and Emergency Operating Procedures," Rev. 3. Chapter 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Rev. 0, July 2000 provides review guidance to the NRC staff for the review of AST amendment requests. Standard Review Plan (SRP) 15.0.1 states that the NRC reviewer should evaluate the proposed change against the guidance in RG 1.183. Chapter 18, "Human Factors Engineering," Rev. 2, provides staff review guidance for planned operator actions.

License Amendment 241, dated October 27, 2004 (ADAMS Accession No. ML042960007),

"Indian Point Nuclear Generating Unit No. 2 - Issuance of Amendment Re: 3.26 Percent Power Uprate," used an alternative source term methodology for analyzing the radiological consequences of seven design-basis accidents using RG 1.183.

The regulatory requirements from which the NRC staff based its acceptance are the reference values in 10 CFR 50.67, and the accident specific guideline values in Regulatory Position 4.4 of RG 1.183 and Table 1 of SRP Section 15.0.1.

10 CFR 50.120, "Training and qualification of nuclear power plant personnel,"

NUREG-1764, "Guidance for the Review of Changes to Human Actions," Rev.1.

Generic Letter 82-33, "Supplement 1 to NUREG-0737- Requirements for Emergency Response Capability,"

NUREG-0700, "Human-System Interface Design Review Guidelines," Revision 2; NUREG-0711, "Human Factors Engineering Program Review Model," Revision 2;

IN 97-78, "Crediting Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times,"

3.0 TECHNICAL EVALUATION

The proposed change would revise TS 3.5.4, "Refueling Water Storage Tank (RWST)," such that the non-seismically qualified piping of the skid-mounted Boric Acid Recovery System (BARS) may be temporarily connected to the RWST seismic piping. Operation of the BARS from the RWST will be under administrative controls for a limited period of time (i.e., 30 days for RWST filtration prior to each fuel cycle). This change is only applicable until Refueling Outage R22 ends (spring 2016). The proposed change would add the following note toTS 3.5.4:

The RWST isolation valves 350, 727 A and 845 connected to non-safety related piping may be opened under administrative controls for up to 30 days per fuel cycle for filtration until the end of refuel outage 22.

This change supports recirculation of the contents of the RWST through the BARS during MODES 1 through 4 for the purpose of silica filtration. This safety evaluation addresses the systems review, the impact of the proposed change on planned operator actions to appropriately maneuver the plant during this configuration, and previously analyzed design basis accident radiological consequences.

3.1 Systems Review During plant operation in Modes 1 through 4, the RWST is required to be operable to maintain a borated water supply for accident mitigation purposes. The RWST is aligned to the suction of the high head safety injection pumps, the residual heat removal pumps, and the containment spray pumps during normal operation in Modes 1 through 4. The suction of the charging pumps is automatically aligned to the RWST on a safety injection signal. During refueling operation in Modes 5 and 6, the RWST is required to be operable as a borated water supply should the boric acid storage system not be operable. The contents of the RWST are also used to flood the refueling cavity during refueling operations. The water in the RWST is borated to a concentration sufficient to ensure shutdown margin is maintained when the reactor is at cold shutdown conditions should RWST water be added to the reactor.

The BARS is a non-seismic skid-mounted unit that is temporarily connected to seismic designed piping for the purpose of removing silica that has accumulated from the gradual deterioration of Boraflex inserts in the Unit 2 spent fuel pool. System operation of the BARS is initiated by opening normally-closed 2-inch manual valves 845 and 727 A that allows RWST water to flow from the 16-inch ECCS header to the refueling water purification pump. A temporary alignment permits the BARS to take suction from manual valve 725 which is located on the discharge side of the refueling water purification pump. The temporary alignment is created by removing the bonnet and internals from valve 725 and installing a hose adapter plate. The piping and manual valves 845, 727 A, and 725 leading to the BARS suction line are all seismically designed. The BARS discharge line is connected to 2-inch manual valve 350 by temporarily removing a 2-inch flange and installing a hose adapter plate. Manual valve 350 is also seismically designed and leads directly back to the RWST.

As described in Section 9.3.2 of the IP2 UFSAR, water from the RWST can be purified by pumping it via the refueling water purification pump to the spent fuel pit demineralizer and filter.

However, processing the RWST contents through the BARS to remove silica as described above isolates the spent fuel pit demineralizer and the spent fuel pool cooling system.

Therefore, the proposed system alignment will not interfere with the seismically designed spent fuel pool cooling system and will not create the possibility to drain the SFP to an unsafe level.

As a result, the NRC staff finds that the proposed system alignment will not impact the SFP or spent fuel pool cooling system as described in the IP2 UFSAR.

The licensee proposes to add a note toTS 3.5.4 to allow the BARS to be aligned to the RWST to filter the RWST water for up to 30 days per fuel cycle until the end of Refueling Outage 22.

This would include both the spring 2014 and 2016 refueling outages. The note would allow this alignment under administrative controls. Administrative controls include procedures to isolate the BARS based on events or direction from the control room. Methods of communication between the operator and the control room are discussed in the procedures.

Valves 845, 727 A, and 350 are seismic class 1 valves and will be part of the lnservice Test Program with a test frequency of two years. By procedure, these valves will by cycled open and closed prior to placing the BARS in operation to provide reasonable assurance that the valves can be closed to isolate the non-seismic BARS if necessary.

The licensee has proposed to assign a dedicated operator (i.e., one who has no other responsibilities during this system alignment) to isolate the non-seismic BARS from the seismic RWST piping by closing the manual operated valves if necessary. The operator will be trained to perform this function and will have the ability to communicate directly with the control room.

Procedures will require the dedicated operator to isolate the BARS following either an RWST low level alarm, a safety injection signal, when lights go out in the primary auxiliary building, a hose connecting either the suction or discharge lines ruptures, or when an indication of tremors or earthquake is apparent. Upon direction, the dedicated operator would ( 1) isolate the BARS suction line by closing valves 845 and 727 A, (2) isolate the BARS discharge line by closing valve 350, and (3) securing the refueling water purification pump if running.

The licensee has calculated the maximum time allowable for operator action to isolate the non-seismic BARS from the seismic RWST piping. The analysis assumes the simultaneous rupture of the BARS suction connection at valve 725 and the BARS discharge connection at valve 350. The analysis assumes simultaneous ruptures with and without an accompanying safety injection signal.

The TS limit of 345,000 gallons of water for the RWST corresponds to an RWST level of 36.83 feet. The current RWST low level alarm is generated when the level reaches 37.01 feet. In order to maximize the time available for the dedicated operator to take action, the licensee has proposed to raise the initial RWST level to a minimum level of 37.43 feet and raise the RWST low level alarm to 37.33 feet prior to placing the BARS into service. For the case where the refueling water purification pump is running and both the BARS suction and discharge connections rupture simultaneously, the licensee's analysis assumes a break flow of 180 gpm at valve 725 and 91 gpm at valve 350. These values are considered to be maximum flow rates based upon pump runout conditions and available RWST head. Under these conditions, the licensee concludes that it would take 31 minutes before the RWST drained down to the minimum allowed TS level. This is the bounding case because the case involving a safety

injection signal would generate an automatic trip of the refueling water purification pump which would limit the respective break flows at the BARS suction and discharge connections.

The licensee performed a verification walkdown to demonstrate the feasibility of control room operators to manually isolate the BARS if necessary. Manual valves 845 and 727 A on the BARS suction side are located within approximately 12 feet of the refueling water purification pump in the primary auxiliary building. Manual valve 350 on the BARS discharge side is also in the primary auxiliary building but at a different elevation. During the verification walkdown, the operator proceeded at a normal pace from the control room and completed the tasks of closing valve 845 and 350 in a total time of 5 minutes. An additional 2 minutes was estimated for tripping the refueling water purification pump and closing valve 727 A resulting in a total time of 7 minutes. This verification walkdown provides reasonable assurance that the seismic RWST will be isolated from the non-seismic BARS within the 31 minutes of time calculated before the RWST water level reaches the TS minimum required level, such that there will be sufficient margin to the TS required RWST level. The estimated operator action time of 7 minutes should reasonably be reduced in an actual seismic event, when considering that a dedicated operator will be present to perform these duties.

Isolation of the non-seismic BARS from the seismic RWST piping can be accomplished by closing manual seismic valves and tripping the refueling water purification pump by a dedicated operator who is trained to perform these duties as necessary. The NRC staff has concluded that there is reasonable assurance that the dedicated operator will be able to perform these required functions within the required 31 minute completion time. As a result, the current TS required RWST water level and the current accident analysis remain valid.

Summary and Conclusion The NRC staff has reviewed the proposed changes toTS 3.5.4 that will permit non-seismically qualified piping of the BARS to be connected to the RWST seismic piping by manual operation of seismically qualified ASME boundary valves under administrative controls for limited time periods. The limited time period is up to 30 days per fuel cycle for filtration of the RWST water.

The changes are to be applied only for the next two fuel cycles. They cannot be used after Refueling Outage R22 (spring 2016). The staff has concluded that there is reasonable assurance that the dedicated operator will be able to perform the required functions within the required 31 minute completion time. Furthermore, the staff has found that there is reasonable assurance that the RWST level will remain above the TS minimum required level in the event of a seismic event. Therefore, staff has concluded that the proposed TS change is acceptable.

3.2 Component Performance and Testing As noted in the licensee's letters dated April15 and September 4, 2013, manual valves 845, 727A and 350 will be added to the Indian Point Unit 2 lnservice Test (1ST) Program as active valves and will be exercise tested in the future according to the applicable requirements of the ASME OM Code of record. The licensee further identified that only manual valves 845 and 727 A will be subject to leakage testing and that this testing will also be performed in the future according to the applicable requirements of the ASME OM Code of record. The Indian Point Unit 2 ASME OM Code of record is currently OM Code 2001 Edition through OMb-2003 Addenda. Under the applicable OM Code requirements and 10 CFR 50.55a regulations, these valves will be exercise tested and leakage tested at a frequency of no longer than two years.

The NRC staff finds that inclusion of the subject valves in the Indian Point Unit 2 1ST Program and proposed testing is sufficient to ensure the operational readiness of these valves.

3.3 Human Performance Description of Operator Action(s) and Assessed Safety Significance The proposed TS change would revise TS 3.5.4, "Refueling Water Storage Tank" to allow the non-seismically qualified piping of the temporary BARS to be connected to, and isolated from, the RWST's seismically qualified piping by manual operation of RWST seismically qualified boundary valves. This action would be performed under administrative controls for limited periods of time, i.e., 30 days per fuel cycle for filtration of suspended solids from the RWST water. This action will be performed by a qualified, dedicated crew member to be available to isolate the RWST in a timely manner, so that the RWST retains its safety function of being a source of reactor coolant during transients and accidents. This is proposed as an interim action until an alternate solution makes the action unnecessary. This change, as proposed, will only be in effect until Refueling Outage R22 (spring 2016) ends.

In accordance with the generic risk categories established in Appendix A to NUREG-1764, this task sequence is considered "risk-important" due to the fact that its failure would jeopardize the ECCS injection and recirculation phases of a loss-of-coolant accident (LOCA). Because of its risk importance, the NRC staff performed a "Level One" review, i.e., the most stringent of the graded reviews possible under the guidance of NUREG-1764.

Operating Experience Review The licensee performed an operating experience review for the proposed manual actions.

Currently, the proposed operator actions have been used in the plant since 2006. There have been no seismic events that would require operators to isolate the RWST from the non-seismic piping, and no failures by operators to properly recirculate the RWST volume in the purification mode. The licensee's search of Condition Reports since 2006 identified only logistic issues such as security clearance for the BARS equipment, manpower scheduling, and tripping hazards. The licensee also correctly identified Farley and Indian Point Unit No.3 as approved precedents for this LAR. Based on IP2's operating history of successful implementation of the proposed manual actions, the NRC staff finds the IP2 operating experience review acceptable.

Functional Requirements Analysis and Function Allocation Because the proposed operator action is not a new action, functional requirements analysis and function allocation are not necessary. Prior experience has shown that operators, when assigned this task, have sufficient time and resources available to perform it reliably. The proposed actions are guided by a controlled procedure. The NRC staff agrees that functional requirements analysis and function allocation are not necessary based on the licensee's long-term use of a controlled procedure, demonstration of adequate margin to proposed time constraints (see Human Factors Verification and Validation, below), their characterization of the action as a time-critical action (see Human Performance Monitoring Strategy, below), and their intent to monitor the feasibility and reliability of the action (see Human Performance Monitoring Strategy, below, for discussion of long-term monitoring) until final resolution is complete, i.e.,

Refueling Outages R22 (spring 2016).

Task Analysis Because the proposed operator manual actions are not new, the only aspect requiring reanalysis is the establishment of time constraints for the action sequence. The licensee established the design value of 31 minutes and an operational value of seven minutes for the time to close the valves that isolate the RWST from the non-seismic BARS. The ability of operators to isolate the RWST within the design and operational values for the timing of the action sequence were validated (see Human Factors Verification and Validation, below). The walkthrough testing demonstrated adequate margin to the operational time limit. The NRC staff finds the Entergy update to the task timing acceptable based on their validation of adequate margin to proposed time constraints.

Staffing Staffing and qualification are not affected by the proposed LAR. No new or additional crew members are required, nor are there any new or additional qualifications required to perform the action sequence within the time constraints established. The licensee stated in its submittal that each crew will have a dedicated operator who is trained on the symptoms of a seismic event and his or her specific responses required to protect the volume of the RWST in accordance with procedure 2-0SP-10.1.1. These actions will only be required when the RWST is connected to non-seismic piping, i.e., 30 days or less per fuel cycle. The NRC staff finds that the licensee's staffing plans for dedicating an operator to manually isolate the RWST acceptable as an interim action until a final resolution is implemented by the end of R22 (spring 2016).

Human-System Interface Design Human-System Interface (HSI) design, including the design of the Safety Parameter Display System (SPDS) will not be affected by the proposed LAR.

Procedure Design No changes are required to the Emergency Operating Procedures (EOPs). Because the proposed actions are not new actions, the existing procedure, procedure 2-0SP-1 0.1.1, does not require any changes. The NRC staff finds the existing procedure acceptable based on ( 1) the staff's confirmation that the required actions are described in the procedure, and (2) the licensee's validation and verification of the procedure to confirm its effectiveness (see Human Performance Monitoring Strategy, below).

Training Program and Simulator Design The IP2 simulator is not capable of modeling the proposed task sequences and will not, therefore, be used in training. A training needs analysis was completed. Based on the fact that the proposed action has a long history of successful implementation, and is supported by written procedures, the licensee concluded that additional training is not necessary. Based on the above, the NRC staff finds that the training already provided is acceptable.

Human Factors Verification and Validation Time testing of the proposed action was performed by the licensee to demonstrate sufficient margin to the established design values. The results of the licensee's study indicate that operators are able to isolate RWST flow to the non-seismic piping of the BARS. The licensee found that the operator action is achievable as directed by procedure, and can be done within seven minutes, providing substantial margin to the 31 minute design value available. Based on the results of these walkthrough demonstrations, the NRC staff finds that the actions are feasible and can be reliably performed by IP2 operators within the calculated time constraint established (31 minutes) using existing, controlled procedures.

Human Performance Monitoring Strategy The procedures and actions proposed by this LAR have been included in the licensee's Time Critical Operator Actions program as controlled by procedure OAP-115. This will ensure that subsequent changes to the plant, procedures, or programs will not invalidate the established action times. Based on the administrative protection against inadvertent change and the cyclical validation required by OAP-115, the NRC staff finds the licensee's long-term monitoring strategy acceptable.

Human Performance Conclusion Based on the statements provided by Entergy, i.e., that time-testing results demonstrate significant margin to design, that appropriate administrative controls will be applied to procedures and training, and that the operators have substantial in-house operating experience, the NRC staff concludes that the proposed LAR is acceptable from the human performance point of view.

3.4 Accident Dose

Background

Per SRP 15. 0.1, the NRC staff evaluated the proposed change against the guidance in RG 1.183. RG 1.183, Regulatory Position 5. 0 states:

ESF [Engineered Safety Features] systems that recirculate sump water outside of the primary containment are assumed to leak. during their intended operation. This release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. This release source may also include leakage through valves isolating interfacing systems (Ref. A-7) [Information Notice 91-56, "Potential Radioactive Leakage to Tank Vented to Atmosphere," dated September 19, 1991]. The radiological consequences from the postulated leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA.

The RG 1.183, Regulatory Position 5.1.2 also states:

Credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed.

After the end of post-LOCA injection phase, the Emergency Core Cooling Systems (ECCS) and containment spray systems [ESF systems that recirculate sump water] are switched to the recirculation phase. This requires that water from the containment sump replace the RWST as the ESF systems water supply. This is accomplished by realigning several system valves that interface between the ECCS and spray systems, and the RWST and the pathways leading back to the RWST. If these system valves are allowed to leak by design or the leakage of these valves is unknown, a leakage path between the containment sump and the RWST and any interfacing piping may exist due to valve seat leakage.

For the design basis LOCA radiological analyses it is assumed that 40% of the core iodine inventory is mixed homogeneously and instantaneously in the primary containment sump water at the time of release from the core. Because the ECCS takes suction from the sump and the sump water is assumed to be radioactive, the consequences of any ECCS leakage are considered in the radiological dose analyses. According to the IP2 Updated Final Safety Analysis Report, Revision 23, 2012, the dose resulting from 4.0 gallons per hour leakage from the ECCS is considered in the IP2 design basis accident analyses.

The LAR dated, April 15, 2013 states:

The RWPP [Refueling Water Purification Pump] will take suction through manual isolation valve 855 [sic- valve 845 per the September 4, 2013 supplement] on line 2 AC-151R#183 which is connected to the 16 inch line from the RWST downstream of isolation valve 846. Normally closed isolation valve 845 will be opened (Drawing A227781, quadrant 1-1 ), and the RWPP will take suction through valve 727 A and discharge to valve 725 (Drawing A227781 quadrants G-1 to E-1 ). Flow through valve 725 adapter plate is to the non-seismic BARS since the spent fuel pit demineralizer is isolated. The flow from valve 725 is through a 2 inch hose adapter plate to the BARS which discharges to seismic line #252 upstream of valve 350. Flow will be through valve 350 and return line 3"-SI-151 R#161 to the RWST. Flow would not be diverted back to the boric acid makeup system due to check valve 294 and normally closed manual valve 295 (see Drawings 9321-F-2736, quadrant E-3 and Drawing 9321-F-2735 (valve 350 only), quadrant 1-4).

The LAR proposes to temporarily use manual actions to isolate the BARS in the event of an actual or potential loss of RWST.

Staff Review The LAR dated, April 15, 2013, provided a technical assessment of the impact of the proposed change on the dose consequences of a LOCA when the BARS is in operation (1 0 CFR 50.67).

The licensee stated the following:

Following the injection phase of a large break LOCA (about 20 minutes) the preferred means of cold leg recirculation is to use the internal recirculation pumps. This results in the fluid being kept inside containment until hot leg recirculation. At about 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, the recirculation pumps send fluid from the containment to the suction of the high head safety injection pumps, with the potential for sump fluid leakage to leak back to the RWST and impact BARS.

This flow path is isolated from the Refueling Water Purification Pump by check valve 847 and motor operated valve 1810 (8"-SI-189R, line#155 on drawing 9321-F-2735). It is possible for any leakage past these valves to migrate to the refueling water purification loop; however, this would be contained as the dedicated operator would close valves 845 and 727 A.

Another potential for sump fluid leakage to impact BARS would be leakage through the two inch Sl mini-flow line back to the RWST that is connected to valve 350. However, this would be limited to leakage through MOV 842/843, which are tested by 2-PT-R048 and have an acceptance criterion of 0.5 gallons per hour (gph). These valves and their acceptance criterion are also governed by the 2.0 gph limit for Emergency Core Cooling System (ECCS) leakage and so there would be no impact on dose.

The licensee's letter dated September 4, 2013, provided the following discussion regarding how the proposed change met Regulatory Position 5.1.2:

" ... a dedicated operator would isolate suction from the RWST by closing valves 845 and 727 A. This pair of valves is seismic 1 and in series and the single failure of one of the valves would be mitigated by the other valve. The dedicated operator would also isolate the return line from the BARS to the RWST by closing valve 350. Any leakage through valve 350 would be limited to leakage past MOV 842/843. This pair of valves is in series and tested with a leakage limit of 0.5 gph, which is accounted for in the radiological analysis."

The licensee stated that external recirculation of ECCS fluids occurs after 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. To show that the valves proposed to be open (845, 350 and 727 A) could be closed before external recirculation occurred (prior to radioactive fluids reaching these valves), the licensee performed a simulation in which an operator dispatched from the control room closed valves 845 and 350 in a total of five minutes. An additional 2 minutes was estimated to trip the purification pump and valve 727 A, resulting in a total of seven minutes. The licensee further stated that this time would be even shorter since there would be a dedicated operator for the BARS.

Using RG 1.183, Regulatory Position 5.1.2, the NRC staff evaluated IP2's proposed credit for manual closure of isolation valve 727 A, 845 and 350. Consistent with Regulatory Position 5.1.2, IP2 stated that valves 727A and 845 are seismic Category 1. Valves 727A and 350 are

required to be operable by TSs (per IP2's TS 5.5.2, "Primary Coolant Sources Outside Containment"). Manual actions are credited for valves 727 A and 845 for a limited duration (30 days per fuel cycle) until the end of refueling outage R22. Per IP2's April 15, 2013 letter, valves 727A, 845 and 350 will be successfully cycled (open-closed) prior to operation of the BARS to provide reasonable assurance that the valves will close. Also, these valves will be added to the lnservice Test Program and classified as Category A. Furthermore, since the time for the operator to manually close the valves isolating the BARS is less than the time for internal recirculation (6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />), there will be no radioactive leakage into the non-seismic piping except the leakage allowed by design for the isolation valves. The licensee stated, and the NRC staff agrees, that a single failure of any of the (in series) valves would be mitigated by the other valve. Any leakage through valve 350 prior to or after its closure will be limited to the leakage past MOVs 842/843 (0.5 gph accounted for in the radiological analysis), and according to the September 4, 2013 letter, these valves will always be closed when the potential for this leakage pathway exists. Based upon the information provided by the licensee and the discussion above, the NRC staff finds that the proposed crediting of the 727 A, 845 and 350 valves for the design basis dose analysis meets the intent of Regulatory Position 5.1.2.

Accident Dose Summary The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological impacts of the proposed TS change. The staff finds that the licensee used analysis methods and assumptions consistent with the analysis provided in the UFSAR. The staff compared the doses estimated by the licensee in the UFSAR to the applicable acceptance criteria. Based upon the licensee's results, the staff finds with reasonable assurance that the licensee's estimates of the total effective dose due to design basis accidents will comply with the requirements of 10 CFR 50.67 and the guidance of RG 1.183.

Accident Dose Conclusion The NRC staff reviewed the analysis used by the licensee to assess the radiological impacts of the proposed changes to TS 3.5.4. For the proposed change the staff finds that the licensee used methods consistent with regulatory requirements and guidance. The staff finds with reasonable assurance that the licensee's estimates of the exclusion area boundary, low-population zone, and control room doses will continue to comply with these criteria.

Therefore, the staff finds the proposed change acceptable with regard to the radiological consequences of postulated design basis accidents.

4.0 STATE CONSULTATION

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

5.0 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 significant hazards consideration, and there has been no public comment on such finding (78 FR 38082). 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.

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Stephanie Galbreath, NRR George Lapinsky, NRR Mark Blumberg, NRR Douglas Pickett, NRR John Billerbeck, NRR Date: December 20, 2013

December 20, 2013 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.2- ISSUANCE OF AMENDMENT RE: CONNECTION OF NON-SEISMIC BORIC ACID RECOVERY SYSTEM TO THE REFUELING WATER STORAGE TANK (TAC NO. MF1440)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 273 to Facility Operating License No. DPR-26 for the Indian Point Nuclear Generating Unit No. 2. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated April 15, 2013, as supplemented by letter dated September 4, 2013.

The amendment revises TS 3.5.4, "Refueling Water Storage Tank (RWST)," to allow for the temporary connection between the non-seismically qualified piping of the Boric Acid Recovery System to the seismically qualified piping of the RWST for the purpose of purifying the contents of the RWST in advance of the spring 2014 refueling outage. Operation in this mode will be under administrative controls and will only be applicable through the end of the spring 2016 refueling outage.

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

Sincerely, Ira/

Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-247

Enclosures:

1. Amendment No. 273 to DPR-26

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsNrrDeEpnb RidsNrrDraAadb JBillerbeck, EPNB LPL 1-1 R/F RidsNrrDirsltsb RidsNrrDraAhpb MBiumberg, AADB RidsNrrDorl RidsRgn 1MaiiCenter RidsNrrDssSrxb SGalbreath, AHPB RidsNrrDorllpl1-1 RidsNrrPMindianPoint RidsNrrDssSbpb JMiller, SRXB RidsNrrLAKGoldstein RidsAcrsAcnw_MaiiCTR GPurciarello, SBPB MHamm,ITSB RidsNrrDoriDpr ADAMS ACCESSION NO.: ML 133126A047 OFFICE LPL 1-1/PM LPL 1-1/LA AHPB/BC AADB/BC(A) SRXB/BC NAME DPickett KGoldstein UShoop by JDozier by CJackson memo dated memo dated DATE 11/25/2013 11/25/2013 10/30/2013 11 I 05 I 2013 12/12/2013 12/20/13 OFFICE SBPBIBC EPNB/BC OGC LPL 1-1/BC NAME GCasto Tlupold STurk BBeasley DATE I I 2013 11 I 20/2013 12/19/2013 12/20/13 OFFICIAL RECORD COPY