Text

Proposed License Amendment Regarding Connection of Non-Seismic Boric Acid Recovery System to the Refueling Water Storage Tank
	ML17104A040
Person / Time
Site:	Indian Point
Issue date:	04/07/2017
From:	Vitale A; Entergy Nuclear Northeast
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-17-035, TAC MF1440
	Download: ML17104A040 (53)
	v • d • e

===-Entergx Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel 914 254 6700 NL-17-035 April 7, 2017 Anthony J Vitale Site Vice President U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Proposed Li~ense Amendment Regarding Connection of Non-Seismic Boric Acid Recovery System to the Refueling Water Storage Tank Indian Point Unit No. 2 Docket No. 50-24 7 License No. DPR-26

REFERENCES:

1)

Entergy Letter NL-13-015, "Proposed License Amendment Regarding Connection of Non Seismic Boric Acid Recovery System to the Refueling Water Storage Tank" (April 15, 2013) (ML13116A007)

2)

NRC Letter to Entergy, "Request for Additional Information Regarding Proposed License Amendment to Temporarily Connect Seismic to Non-Seismic Piping under Administrative Controls" (TAC NO.

MF1440), (August?, 2013) (ML13207A387)

3)

Entergy Letter NL-13-115, "Response to Request for Additional Information Regarding Proposed License Amendment to Temporarily Connect Seismic to Non-seismic Piping under Administrative Controls" (TAC No. MF1440), (September4, 2013) (ML13253A138)

4)

NRC Letter, "Indian Point Nuclear Generating Unit No. 2 - Issuance of Amendment Re: Connection of Non-Seismic Boric Acid Recovery \\

System to the Refueling Water Sforage Tank" (TAC No. MF1440)

(December 20, 2013) (ML133126A047)

5)

NRC Letter, "Indian Point Nuclear Generating Unit No. 2 - Correction Letter to Amendment No. 273 Re: Connection of Non-Seismic Boric Acid Recovery System to the Refueling Water Storage Tank" (TAC No. MF1440) (January 9, 2014) (ML14002A431)

6)

Entergy Letter NL-17-021, "Notification of Permanent Cessation of Power Operations, Indian Point Nuclear Generating Unit Nos. 2 and 3" (February 8, 2017)

Dear Sir or Madam:

In Reference 1, Entergy Nuclear Operations, Inc., (Entergy) requested a License Amendment to Operating License DPR-26, Docket No. 50-247 for Indian Roint Nuclear Generating Unit No.

2 (IP2). The amendment revised Technical Specifications (TS) 3.5.4, "Refueling Water

(~

NL-17-035 Docket 50-247/

Page 2 of 3 Storage Tank (RWST}," to allow for the temporary connection between the non-seismically qualified piping of the Boric: Acid Recovery System (BARS) 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. The request stated that 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.

The Commission issued Amendment No. 273 (References 4 and 5), which consisted of changes to the TS in response to Reference 1, supplemented by Reference 3 (also attached as Enclosure 2) in response to the NRC Request for Additional Information (Reference 2).

Entergy had planned to install modifications to the BARS piping in order to qualify them seismically prior to the IP2 Spring 2018 Refueling Outage (2R23). However, due to the permanent cessation of IP2 power operation, as requested in Reference 6, and that the 2R23 refueling outage will be the final IP2 refueling outage, there will be limited benefits for the implementation of the planned modifications, considering the required effort.

Pursuant to 10 CFR 50.90, Entergy hereby requests a License Amendment to Operating License DPR-26, Docket No. 50-247 for IP2. The proposed TS change contained herein would revise 3.5.4, "Refueling Water Storage Tank (RWST)" such that the non-seismically qualified piping of BARS be connected to the RWST seismic piping. As in 2R22, operation of the BARS from the RWST will be under administrative controls for a limited period of time (i.e., 30 days for removal of silica from the RWST water). This change will only be applicable until the end of IP2 Refueling Outage 2R23.

v Please note that a similar request was asked for and granted by the NRC for Unit 3 for operation using this Relief up until the end of 3R18. No such change for Unit 3 will be requested by Entergy.

Entergy has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1) using the criteria of 10 CFR S0.92(c) and has determined that this proposed change involves no significant hazards, as described in Attachment 1. The marked up page showing the proposed change is provided in Attachment 2. The associated Bases change is provided in Attachment 3 for information. A copy of this application and the associated attachments are being submitted to the designated New York State official in accordance with 1 O CFR 50.91.

Entergy requests approval of the proposed amendment by January 20, 2018 and an allowance of 30 days for implementation. There are no new commitments being made in this submittal. If you have any questions or require additional information, please contact Mr. Robert Walpole, Manager, Licensing at (914) 254-6710.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April

--=f. ' 2017.

Sincerely, AJV/mm

NL-J?-035 Docket 50-247 Page 3 of 3 Attachments:

1.

Analysis of Proposed Technical Specifications Change Regarding

Enclosure:

Connection of Non Seismic BARS to Refueling Water Storage Tank

2.

Marked Up Technical Specifications Page for Proposed Change I

Regarding Connection of Non Seismic BARS to Refueling Water Storage Tank

3.

Marked Up Technical Specifications Bases Change Associated with the Proposed Change Regarding Connection of Non Seismic BARS to Refueling Water Storage Tank

1.

Indian Point 2 Drawings and Calculation 2

Entergy Letter NL-13-115, "Response to Request for Additional Information Regarding Proposed License Arnendment to Temporarily Connect Seismic to Non-seismic Piping under Administrative Controls" (TAC NO. MF1440}, (September4, 2013) (ML13253A138) cc:

Mr. Daniel H. Dorman, Regional Administrator. Region I, NRC Mr. Douglass Pickett, Senior Project Manager, NRR/DORL, NRC Ms. Bridget Frymire, New York State Department of Public ~ervice Mr. John B. Rhodes, President and CEO NYSERDA NRC Resident Inspector's Office

ATTACHMENT 1 TO NL-17-035 ANALYSIS,OF PROPOSED TECHNICAL SPECIFICATIONS CHANGE REGARDING CONNECTION OF NON SEISMIC BARS TO REFUELING WATER STORAGE TANK ENTERGY NUCLEAR OPERATIONS, INC.

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

1.0 DESCRIPTION

{

I NL-17-035 Docket No. 50-247 Page 1of8 Entergy Nuclear Operations, Inc. (Entergy) is requesting an amendment to Operating License DPR-26, Docket Nb. 50-247 for Indian Point Nuclear Generating Unit No. 2 (IP2). The proposed Technical Specifications (TS) change contained herein would revise 3.5.4, "Refueling Water Storage Tank (RWST)" such that the non-seismically qualified piping of the Boric Acid Recovery System (BARS) may be connected to RWST seismic piping, and isolated by manual operation of RWST seismically qualified boundary valves under administrative controls for a limited period of time (Le., 30 days for filtration for removal of silica from the RWST water). This change will be applicable for the next and last IP2 Refueling Out~ge 2R23 (Spring 2018). If Unit 2 operates past 2020, Entergy will address this issue either through a modification or water processing. Entergy will not ask for additional relief.

The specific proposed change is listed in the following section.

2.0 PROPOSED CHANGE

The proposed TS change is as follows:

Directly unde.r "LCO" add

- NOTE-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 for filtration until the end of Refueling' Outage 2R23.

In addition, the Technical Specifications Bases will be revised to clarify this issue.

3.0 BACKGROUND

Historically, IP2 was connebting the non-seismic reverse osmosis system, identified as the: BARS to the seismic Spent Fuel Pool (SFP) Purification Loop to filter RWST water while in plant conditions and modes for which the RWST was required to be operable. This alignment was utilized to remove silica from the RWST water. Removal of silica is necessary to maintain Reactor Coolant System (RCS) chemistry within fuel requirements and to improve water clarity during refueling to facilitate safe handling of fuel and to prevent delays in fuel movement. The water clarity is both a personnel and equipment safety consideration. Entergy had established the practice of recirculating the RWST for up to 30 days beginning up to about two months prior to a refueling outage for silica removal. Prior to Refueling Outage (RO) 2R21 the RWST was recirculated for a duration of 11 days. After recirculation the total concentration of silica was 1.9 ppm. Prior to RO 2R22, the RWST was recirculated for a duration of 15 days. A sample taken after recirculation had total concentration of silica of 1.05 ppm.

NL-17-035 Docket No. 50-247 Page 2 of 8 During plant operations 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 (Modes 1 through 4). During cold shutdown and refueling operation (Modes 5 and 6), the RWST may be credited as a borated water supply should the boric acid storage system not be functional. The contents of the RWST are also used to flood the refueling cavity during refueling operation. The water in the RWST is borated to a concentration sufficient to ensure that shutdown margin is maintained when the reactor is at cold shutdown conditions should RWST water be added to the reactor.

It was recognized that alignment to BARS could render the RWST inoperable during a seismic event since the BARS is a non-seismic system. To maintain operability, procedure changes had been made to direct manual operator action to isolate the non-seismic connections to the RWST to ensure adequate inventory during Modes 1 through 4 when the RWST was required to be operable. After reviewing Information Notice (IN) 2012-01, "Seismic Considerations-Principally Issues Involving Tanks," Entergy concluded that manual actions could not be credited for this purpose without prior NRC approval and subsequently discontinued this practice. The SFP Purification Loop is a subsystem of the spent fuel pool cooling system that is connected to portions of the RWST piping. The SFP Purification Loop piping has been upgraded to seismic Class '1 so that during a seismic event no failure of the SFP Purification Loop piping is considered. However, when the non-seismic BARS is connected to the Purification Loop, there is the possibility that a seismic event could affect the available water in the RWST. For this reason the IN 2012-01 requires that the RWST TS action statement be entered when non-seismic connections are made to the RWST. The completion time of the action statement does not allow time for purification.

Removal of silica by use of the BARS system is preferred to other means. For example, using dilution creates large quantities of liquid radioactive waste, or removing silica from the spent fuel pool has the potential for further deterioration of the Boraflex material in the storage racks.

Consequently, this TS change request is being made to credit operator action to close the seismically qualified manual code boundary valves in the event of a seismic or design basis accident.

4.0 Technical Evaluation This assessment addresses the proposed change to TS 3.5.4, "Refueling Water Storage Tank (RWST)." The TS change would allow voluntary connection of the non-seismic BARS to the seismic piping of the SFP Purification System with a 30 day limit for re-circulating the*contents of the RWST for the purpose of silica filtration through the BARS during Modes 1 through 4 when the RWST is required to be operable. The following assessment provides the basis for the acceptability of the proposed change to the TS which provides for operator action to close the seismically qualified manual code boundary valves to assure RWST operability when re-circulating the tank through non-safety related piping.

The non-seismic BARS is connected to the seismic SFP Purification System as follows:

The BARS suction line is connected to Valve 725 (see Drawing A227781, quadrant F-1 in the Enclosure) on the discharge to the Refueling Water Purification Pump (RWPP) by removing the valve bonnet and valve internals and installing a hose adapter plate.

NL-17-035 Docket No. 50-247 Page 3 of 8 The BARS discharge line is connected to 2 inch line #252 upstream of 2 inch valve 350 (see Drawing 9321-F-2736, quadrant E-3 and Drawing 9321-F-2735 (valve 350 only),

quadrant 1-4 in the Enclosure) by removing the 2"-150 psig flange and installing a hose adapter plate. '

\\

The RWPP will take suction through manual isolation valve 845 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 727A 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 #253 upstream of valve 350. Flow will be through valve'-

350 and return line 3"-Sl-151R#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 proposed manual action to isolate the BARS in the event of an actual or potential loss of RWST considered the following:

Operator Action Considerations l

~

Entergy has confidence in the successful completion of manual actions due to the training program completed for all system operators and the specific procedural requirements for the BARS. During use of the BARS, the RWST level, temperature and boron concentration are monitored. A dedicated operator is assigned to remain in the vicinity of BARS at all times when the RWST Silica Cleanup Skid is aligned for operation. The operator has the ability to directly communicate with the IP2 control room, is equipped with an operational flashlight, and is trained on the location and operation of valves and the Refueling Water Purification Pump (Refere.nce 2). If there is a RWST low level alarm received the Unit 2 control room supervisor will direct the operator to isolate the RWST Silica Cleanup System. The RWST Silica Cleanup System would also be isolated if:

There is a Safety Injection (SI) actuation Lights go out in the PAB A RWST Silica Cleanup System Hose ruptures or breaks An indication of tremors or earthquake is evident If the BARS has to be isolated for any of the reasons above, the dedicated operator would isolate suction to,BARS through valves 845 and 727 A, isolate discharge from BARS through valve 350, and secure the RWST Purification pump, if running.

Valves 845, 727A and 350 will be part of the In-service Test Program 'with a test frequency of two years. Further, by procedure, valves 845, 727 A and 350 will be cycled open and closed prior to putting the BARS in operation to provide reasonable assurance that all valves will close.

I The allowable time for operator actiori to isolate the BARS unit has been calculated (Reference 1). This re-analysis was conservatively based on a simultaneous rupture of connections at valve 725 and at the flange upstream of valve 350. Scenarios for rupture with and without an SI signal were evaluated. The current RWST low level alarm is set to 37.01 feet, with the TS limit at 36.83 feet. In order to provide more time for the operator to perform the isolation function in a seismic or SI event during operation of the BARS, the initial'*level of the RWST

NL-17-035 Docket No. 50-247 Page 4 of 8 would be raised to 37.43 feet or higher, and a control room Plant Integrated Control System (PICS) alarm setpoint would be set at 37.33 ft (or higher) prior to aligning the BARS. There would then be 4684 gallons of margin to the Technical Specifications limit (345,000 gallons) following a low RWST level alarm. If the RWST purification pump (RWPP) is in service providing flow to the BARS unit, a flow of 180 gallons per minute (gpm) was considered for the break flow through valve 725 and 91 gpm was considered for the break flow through the flange at valve 350. These are maximum flow rates resulting from pump runout and available head in the RWST. The total time available before reaching the TS limit would be 31 minutes, assuming the operator took 10 minutes to close valves 845/727 A and trip the RWPP (all in close proximity), and an additional 21 minutes to close valve 350.

For the same break(s) scenario, and considering actuation of SI, the RWPP would receive a trip signal, and the corresponding total flow through the two break locations would be less limiting than the above scenario with no SI signal.

The refueling water purification pump is located on the 68 foot elevation of the Primary Auxiliary Building (PAB) with the pump control switch on an adjacent wall. Valves 845 and 727A are within about twelve feet of each other on opposite sides of the pump. The return line isolation valve 350 is located on the PAB 98.0 foot elevation. There is a card reader at the entry point to the PAB on the 80 foot elevation, but once inside there are no restrictions to reach valve 350 from valves 845 and 727 A. A simulation performed by Operations, with an operator dispatched from the control room resulted in closure of valves 845 and 350 in a total of 5 minutes. An additional 2 minutes is conservatively estimated for tripping the RWPP and closing valve 727 A, resulting in a total of 7 minutes. This time would be even shorter since there would be a dedicated operator for the BARS.

\\ This provides substantial margin to the total calculated time of 31 minutes to shutc;lown the BARS and maintain the RWST within the TS value following a control room alarm indication of 37.33 feet for RWST level.

Dose consequences associated with the operation of the BARS The dose consequences in the highly unlikely event of a Loss of Coolant Accident (LOCA) when BARS is in operation are discussed~below.

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 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , 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 the BARS. This flow path is isolated from the RWPP by check valve 847 and motor operated valve 1810 (8"-Sl-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 the BARS would be leakage through the 2 inch SI 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 surveillance tested by 2-PT-R048 and have an acceptance criterion of 0.5 gallons per hour (gph). These valves and their acceptance criteria are also governed by the 2.0 gph limit for Emergency Core Cooling System (ECCS) leakage, so there would be no impact on dose.

NL-17-035 Docket No. 50-247 Page 5 of 8 The IP2 design is fairly unique in having internal recirculation pumps as well as residual heat removal (RHR) pumps. Use of the RHR is the secondary means of achieving hot leg recirculation by drawing water from the containment sump and delivering via the RWST suction line. This leakage pathway is not postulated because it would require a passive failure, which is not postulated to occur for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The leakage associated with this pathway is not part of the TS 5.5.2 program, because that program does not assume the single passive failure. Likewise, the Regulatory Guide (RG) 1.183 guidance does not impose any additional single failure to determine this leakage path.

Thus, in the highly unlikely event of a LOCA during the operation of the BARS, there will be no impact on accident dose consequences.

Operating Experience The BARS has been in -use at IP2 since prior to refueling outage 2R16 in 2006, and Reference 1 captures operator actions for isolation of BARS for any of the conditions discussed above. A search of Condition Reports since 2006 identified only logistic issues such as security clearance of the BARS equipment, manpower scheduling, tripping hazard due to BARS hoses, etc. There have been no seismic events during the use of the BARS and no problems identified in the installation, use and removal of BARS.

~

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) has evaluated the safety significance of the proposed change to the Indian Point 2 Technical Specifications (TS) which revise TS 3.5.4, "Refueling Water Storage Tank (RWST}," to allow administrative control of the seismic RWST/non-seismic BARS interface.' The proposed change has been evaluated according to the criteria of 1 O CFR 50.92, "Issuance of Amendment". Entergy has determined that the subject change does not involve a Significant Hazards, Consideration, as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The use of the non-seismic Boric Acid Recovery System (BARS) to re-circulate and filter the RWSTwater does not involve any changes or create any new interfaces with the reactor coolant system or main steam system piping. Therefore, the connection of the BARS Purification Loop to the RWST would not affect the probability of these accidents occurring. The BARS is not credited for safe shutdown of the plant or accident mitigation. Administrative controls ensure that the BARS can be isolated as necessary and in sufficient time to assure that the RWST volume will be adequate to perform the safety function as designed. Since the RWST will continue to perform its safety function and overall system performance is not affected, the consequences of the accident are not increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

I

NL;.17-035 Docket No. 50-247 Page 6 of 8

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. The design of the RWST and the SFP Purification Loop has been revised to allow recirculation and purification using the BARS for a short period of time (not to exceed 30 days per fuel cycle) for the next fuel cycle. The BARS takes RWST water in and processes it out without additional connections that could affect other systems and without an impact from its installation. Procedures for the operation of the plant, including the BARs, will not create the possibility of a new or different type of accident. Contingent upon manual operator action, a BARS line break will not result in a loss of the RWST safety function. Similarly, an active or passive failure in the BARS will not affect safety related structures, systems or components.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No. The SFP Purification Loop and recirculation and purification of the RWST water using the BARS is not credited for safe shutdown of the plant or accident mitigation.

RWST volume will be maximized prior to purification and timely operator action can be taken to isolate the non-seismic system from the RWST to assure it can perform its function. This will result in no significant reduction in the margin of safety~

Therefore the proposed change does not significantly reduce the margin of safety.

Based on the above, Entergy concludes that the proposed amendment to the IP2 TS presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of 'no significant hazards consideration' is justified.

5.2 Applicable Regulatory Requirements/ Criteria The NRC Order of February 11, 1980 required an evaluation of the degree of compliance with the GDC at the time. This section discusses continued compliance with certain of those criteria.

The plant will continue to meet Criterion 1 of 10 CFR 50 Appendix A which says "Structures, systems and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems and components will satisfactorily perform their safety functions.

. Appropriate records of the design, fabrication, erection, and testing of structures, systems and components important to safety shall be maintained by or under the control of the nuclear power plant licensee throughout the life of the unit" and Criterion 2 which says "Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability

NL-17-035 Docket No. 50-247 Page 7 of 8 to perform their safety functions. The design bases for these structures, systems and components shall reflect: ( 1) appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accur~cy, quantity, and period of time in which the historical data have been accumulated, (2) appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed."

The purification of the RWST will use the seismic piping meeting these criteria but will also use the non-seismic piping which does not. Manual action will be used until the end of the next two refuel outages to assure isolation of the seismic piping from the non-seismic piping during any condition requiring the RWST volume to be intact and threatening to reduce the RWST level below the TS allowable. This will assure continued compliance with these criteria.

The plant will continue to meet Criterion 35 which says "A system to provide abundant emergency core cooling shall be provided. 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." The RWST provides a support function for this criterion since it supplies the water which is injected following an event and must contain the amount of water required by analysis. Manual action will be used until the end of the next refuel outage to assure isolation of the seismic piping from the non-credited non-seismic piping to assure RWST level meets the TS allowable. This will assure continued compliance with this criterion.

5.3 Environmental Considerations The proposed changes to the IP2 TS do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation

exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 PRECEDENCE Joseph M. Farley Units 1 and 2 received approval for taking manual action to isolate the RWST from the non-seismic SFP lines in Amendments 188 and 183, respectively (Reference 3).

Indian Point 3 received approval f_or taking manual action to isolate the RWST from the non-seismic SFP* lines in Amendment 250 (Reference 4).

Indian Point 2 received approval to isolate the RWST from the non-seismic SFP lines in Amendment 273 (Reference 5 and 6).

7.0 REFERENCES

1. 2-0SP-10.1.1, Support Procedure - Safety Injection Accumulators and Refueling Water Storage Tank Operations.

NL-17-035 Docket No. 50-247 Page 8 of 8

2. IP-CALC-13-00005, Rev 1, "Engineering Evaluation of Postulated RWST Inventory Loss During the Reverse Osmosis Clean-up Skid Process in Accordance to 2-TAP-001-ROS due to a seismic Event", March 2013.

3. NRC Letter to Southern Nuclear Operating Company, Inc., "Joseph M. Farley Nuclear Plant, Units 1 and 2, Issuance of Amendments regarding Refueling Water Storage Tank (TAC NOS. ME8005 AND ME8006)", dated March 24, 2012.

~

4. NRC letter to Entergy,,"Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Connecting Non-Seismic Purification System Piping to the Refueling Water Storage Tank (TAC NO. ME9263)" (February 22, 2013) (ML13046A166)

5. NRC Letter, "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) (December 20, 2013) (ML 133126A047)

6. NRC Letter, "Indian Point Nuclear Generating Unit No. 2 - Correction Letter to Amendment No. 273 Re: Connection of Non-Seismic Boric Acid Recovery System to the Refueling Water Storage Tank" (TAC No. MF1440) (January 9, 2014) (ML14002A431)

/

ATTACHMENT 2 TO NL-17-035 MARKED UP TECHNICAL SPECIFICATIONS PAGE FOR PROPOSED CHANGE REGARDING CONNECTION OF NON SEISMIC BARS TO REFUELING WATER STORAGE TANK Unit 2 Affected Page:

3.5.4-1 ENTERGY NUCLEAR OPERATIONS, INC.

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

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-RWST 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 fRefueling aOutage

~2R23.

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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months reasons other than OPERABLE status.

Condition A or B.

D.

Required Action and D.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND

NOTE -------------------

LCO 3.0.4.a is not applicable when entering MODE 4.

D.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months INDIAN POINT 2 3.5.4 - 1 Amendment No.

ATTACHMENT 3 TO NL-17-035 MARKED UP TECHNICAL SPECIFICATION BASES CHANGE ASSOCIATED WITH THE PROPOSED CHANGES REGARDING CONNECTION OF NON SEISMIC BARS TO REFUELING WATER STORAGE TANK Unit 2 Affected Page:

3.5.4-4 ENTERGY NUCLEAR OPERATIONS, INC.

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

.BASES RWST B 3.5.4 APPLICABLE SAFETY ANALYSES (continued)

LCO INDIAN POINT 2 the low low alarm setpoint and sufficient coolant inventory to support pump operation in recirculation mode is verified to be in the containment.

The RWST level low low alarm setpoint has both upper and lower limits. The upper limit is set to ensure that switchover does not occur until there is adequate water inventory in the containment to provide ECCS pump suction.

(This is confirmed by recirculation and/or containment sump level indication.)

The lower limit is set to ensure switchover occurs before the RWST empties, to prevent ECCS pump damage.

1 Requiring 2 channels of RWST level low low alarm ensures that the alarm function will be available assuming a single failure of one channel.

The RWST satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical, following a DBA, and to ensure adequate level in the recirculation and containment sump to support ECCS operation in the recirculation mode.

To be considered OPERABLE, the RWST must meet the water volume, boron concentration, and temperature limits establish~d in the SRs.

RWST OPERABILITY requires OPERABILITY of two channels of the RWST level low. low alarm. This is required because the IP2 ESFAS design does not include automatic switchover from the safety injection mode. to the recirculation mode of operation based on low level in the RWST coincident with a safety injection signal. This function is performed manually by the operator who must be alerted by redundant alarms that annunciate RWST level low low. The switchover to the cold leg recirculation phase is manually initiated when the RWST level has reached the low low alarm setpoint and sufficient coolant inventory to support pump operation in recirculation mode is verified to be in the containment.

A note allows the RWST valves that isolate non-seismic piping to be opened under administrative control for filtration until the end of Refueling Outage

2R23.RO 22..

B 3.5.4-4 Revision~

ENCLOSURE,1 TO NL-17-035 INDIAN POINT 2 DRAWINGS AND CALCULATION Unit 2 Documents:

Drawing 227781 Drawing 9321-F-2735

', Drawing 9321-F-2736 Calculation IP-CALC-13-00005, Rev 1 ENTERGY NUCLEAR OPERATIONS, INC.

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ENCLOSURE 2 TO NL-17-035 Entergy Letter NL-13-115, "Response to Request for Additional Information Regarding Proposed License Amendment to Temporarily Connect Seismic to Non-seismic Piping under Administrative Controls" (TAC NO. MF1440), (September 4, 2013) (ML13253A138)

ENTERGY NUCLEAR OPERATIONS, INC.

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

~Entergx NL-13-115 September 4, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852

Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box249 Buchanan, NY 10511-0249 Tel 914 254 6700 John A Ventosa Site Vice President Administration

SUBJECT:

Response to Request for Additional Information Regarding Proposed License Amendment to Temporarily Connect Seismic to Non-seismic Piping under Administrative Controls (TAC NO. MF1440)

Indian Point Unit Number 2 Docket No. 50-247 License No. DPR-26

REFERENCES:

1.

Entergy Letter NL-13-015 to NRC, Proposed License Amendment Regarding Connection of Non Seismic Boric Acid Recovery System to the Refueling Water Storage Tank, dated April 15, 2013

2.

NRC Letter to Entergy, Request for Additional Information Regarding Proposed License Amendment to Temporarily Connect Seismic to Non-Seismic Piping under Administrative Controls (TAC NO. MF1440), dated August?,2013

Dear Sir or Madam:

Entergy Nuclear Operations, Inc (Entergy) requested a License Amendment, Reference'1, for Indian Point Nuclear Generating Unit No. 2 (IP2). The proposed amendment would revise Technical Specification 3.5.4, to allow connection of the non-seismically qualified piping of the temporary Boric Acid Recovery System to the Refueling Water Storage Tank under administrative controls for a limited period of time. On August 7, 2013, the NRC staff identified the need for additional information to complete their review (Reference 2). Entergy is providing additional information in response to this request in Attachment 1 and Enclosure 1.

NL-13-115 Docket No. 50-24 7 Page 2 of 2 A copy of this response is being submitted to the designated New York State official in accordance with 10 CFR 50.91.

There are no new commitments being made in this submittal. If you have any questions or require' additional information, please contact Mr. Robert Walpole, Manager, Licensing at (914) 254-6710.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September

_!::L, 2013.

Sincerely, JAV/ai

Attachment:

Enclosure:

1. Response to Request for Additional Information Regarding Proposed License Amendment to Temporarily Connect Seismic to Non-Seismic Piping under Administrative Controls

1. Indian Point Calculation IP-CALC-11-00091, AST Analysis of IP2 to address the impact of Containment sump solution back-leakage to the RWST after LOCA cc:

Mr. Douglas Pickett, Senior Project Manager, NRC NRR DORL Mr. William Dean, Regional Administrator, NRC Region 1 NRC Resident Inspector Office Mr. Francis J. Murray, Jr., President and CEO, NYSERDA Ms. Bridget Frymire, New York State Dept. of Public Service

ATTACHMENT 1 TO NL-13-115 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED LICENSE AMENDMENT TO TEMPORARILY CONNECT SEISMIC TO NON-SEISMIC PIPING UNDER ADMINISTRATIVE CONTROLS ENTERGY NUCLEAR OPERATIONS, INC.

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

Response To Request For Additi.onal Information Accident Dose Branch Questions and Responses Question 1 NL-13-115 Docket No. 50-247 Page 1of10 Final Safety Analysis Report (FSAR) Section 14.3.6.6, "External Recirculation," provides a description of the analyses used to justify the proposed change (2.0 gallon per hour limit for

- Emergency Core Cooling System (ECCS) leakage).

FSAR Section 14.3.6.6 states:

Since the leakage is initiated at 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the LOCA [loss of coolant accident],

it does not contribute to the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary dose [exclusion area boundary dose or EAB].

Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," states:

The methodology and assumptions for calculating the radiological consequences should reflect the regulatory positions of RG-1.183 [Regulatory Guide 1.183].

Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Regulatory Position 4.1.5, states:

The TEDE should be determined for the most limiting person at the EAB. The maximum EAB TEDE for any two-hour period following the start of the radioactivity release should be determined and used in determining compliance with the dose criteria in 1 O CFR50. 67. 14 The maximum two-hour TEDE should be determined by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successiv,e two-hour periods. The maximum TEDE obtained is submitted. The time increments should appropriately reflect the progression of the accident to capture the peak dose inteNal between the start of the event and the end of radioactivity release (see also Table 6).

This is consistent with Title 10 of the Code of Federal Regulations [ 10 CFRJ, Section 50.67, "Accident Source Term," that states:

An individual located at any point on the boundary of the exclusion area for any

[emphasis added] 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)2 total effective dose equivalent (TEDE). '*

a) Please confirm whether the dose due to ECCS leakage is excluded from the FSAR Section 14.3.6.6 EAB dose calculation.

NL-13-115 Docket No. 50-247 Page 2of10 b) If so, please explain how this is consistent with 10 CFR 50.67. SRP 15.0.1and10 CFR 50.67 both state that the worst dose for any 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period is to be used to determine the EAB dose. This would typically mean the ECCS dose should be added to the time dependent EAB dose and the worst 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose should be determined from this time dependent dose profile. Please justify why the ECCS leakage is not considered in the determination of the EAB dose, or include the ECCS leakage in the EAB dose calculation.

Response to Question 1 a) Any ECCS leakage for the first 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months following a LOCA is internal-to the containment and inherently accounted for in the offsite dose contribution for containment leakage. In order to identify the worst two hour period, the computer runs included time steps to provide EAB 2-hour doses at 0.2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months intervals. As shown below, the worst two-hour dose is 16.91 rem over the 0.6 to 2.6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months interval (this dose was increased by a factor of 1.05 for conservatism and rounded to 17.8 rem as reported in FSAR Section 14.3.6.8). The dose gets reduced to16.47 rem in the 0.8 to 2.8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, and further reduced in the 1.0 to 3.0 hour0 days 0 hours 0 weeks 0 months period. After 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , when ECCS leakage begins outside of containment, the EAB dose. rate from containment airborne leakage is so low that the added radiological contribution from the ECCS leakage pathway is not sufficient to change the maximum 2-hour dose from the peak value set earlier in the accident.

Exclusion Area Bounda Dose rem TEDE 0.4-2.4 hr 0.6 - 2.6 hr 0.8 - 2.8 hr 1.0-3.0hr 16.59 16.91 16.47 15.36 b) See response to a) above.

\\

Question 2 UFSAR Section 14.3.6.6 states:

The releases would be subject to filtration by the filtered ventilation system provided for the primary auxiliary building which houses the portions of the EGGS located

. outside containment.

However, filtration of the releases is not credited in the analysis.

a) Are releases from non-seismic piping (postulated to fail) subject to the filtered ventilation system in the primary auxiliary building?

Response to Question 2 a) Any break in non-seismic piping in the primary auxiliary building would be subject to the filtered ventilation system.

Question 3 NL-13-115 Docket No. 50-247 Page 3of10 The Nuclear Regulatory Commission's safety evaluation, which reviewed the conversion to 10 CFR 50.67, reviewed an analysis which appears to have different assumptions than those provided in FSAR Section 14.3.6.6.

a) Has the NRC staff reviewed the analysis provided in FSAR Section 14.3.6.6 or were these changes made using 10 CFR 50.59, "Changes, tests and experiments"? If a staff evaluation of this analysis has not been performed, please provide the inputs, assumptions, methodology and results of t_he analysis that is to be used to support the proposed change.

b) FSAR Section 14.3.6.6 provides design basis dose values for two different assumptions (assuming a boundary layer effect and assuming no boundary layer effect). Which assumption is used for the licensing basis calculation?

Response to Question 3 a) FSAR Section 14.3.6.6 was revised using 10 CFR 50.59, "Changes, tests and experiments", to include potential ECCS back-leakage to the RWST. A copy of the calculation used to support the change is provided in Enclosure 1 as requested.

b) The licensing basis calculation is based on no boundary layer effect resulting in a Control Room Dose of 4.9 rem. This was reviewed and approved by the NRC in the Safety Evaluation for SPU (NRC Letter to Entergy, Indian Point Nuclear Generating Unit No. 2-Issuance of Amendment Re: 3.26 Percent Power Uprate (TAC NO. MC1865),

October 27, 2004).

Question 4 Page 3 of 8 of the submittal states:

The RWPP [Refueling Water Purification Pump] will take suction through manual isolation valve 855 on line...

a) Please confirm whether this sentence should state valve 845 or whether valve 855 is correct.

Response to Question 4 a) The sentence on page 3 of 8 of the submittal contains a typographical error and should state:

The RWPP [Refueling Water Purification Pump] will take suction through manual isolation valve 845 on line...

NL-13-115 Docket No. 50-247 Page 4of10 Question 5 RG 1.183, Regulatory Position 5.1.2 states:

_ 5.1.2 Credit for Engineered Safeguard Features 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 expJ;cit/y addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences.

a) Please describe how the valves credited to isolate the non-seismic pathways after a design basis accident meet the above regulatory position. For those valves that do not meet the regulatory position please explain the differences between the design features, analytical techniques and procedural methods proposed and the regulatory position and

_ justify how the proposed alternatives to the regulatory position proved an acceptable method for complying with the NRC regulations (10 CFR 50.67).

Response to Question 5 a) As noted in the submittal, a dedicated operator would isolate suction from the RWST to BARS by closing valves 845.and 727A. 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.

Question 6 Page 4 of 8 of the submittal states:

Another potential for sump fluid leakage to impact BARS would be leakage through the' 2 inch SI mini-flow line back to the RWST that is connected to valve 350.

However, this would be limited to leakage through MOV 8421843, which are tested by 2-PT-R048 and have an acceptance criterion of 0.5 gallons per hour (gph).

a) Are MOV 842/843 always closed when the potential for this leakage pathway exists? If not, explain the timing involved for closing MOV 842/843 and valve 350. Can the NL-13-115 Docket No. 50-247 Page 5of10 timing of the closure of these valves cause the 0.5 gph leakage limit to the non-seismic piping to be exceeded for any time period after the start of the postulated accident.

Response to Question 6 a) MOV 842/843 would always be closed when the potential for this leakage pathway exists. For hot leg recirculation, Procedure 2-ES-1.4, '.'Transfer to Hot Leg Recirculation", requires SI pump mini-flow valves MOV-842/843 to be closed. Similarly, for cold leg recirculation with the SI pumps taking suction from the recirculation pumps, 2-ES-1.3, "Transfer to Cold leg Recirculation", requires verifying MOV-842/843 are closed.

Question 7 Page 4 of 8 of the submittal states:

Following the injection phase of a large break LOCA (about 20 minutes) the preferred [emphasis added] 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 6.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months s].

RG 1.183, Regulatory Position 5.1.3 states:

The numeric values that are chosen as inputs to the analyses required by 10 CFR

50. 67 should be selected with the objective of determining a conservative postulated dose.

a) Confirm that plant procedures do not allow the recirculation of sump fluids outside containment prior to 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

b) If plant procedures do allow the recirculation of sump fluids outside of containment prior to 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months why aren't these methods of recirculation considered in the determination of the ECCS leakage dose *calculation?

c) RG 1.183, Regulatory Position 1.3 defines the scope of required analyses which include post accident access shielding (NUREG-0737, "Clarification of TMI Action Plan Requirements," Action Item 11.8.2, "Post-Accident Access Shielding"). If plant procedures do allow the recirculation of sump fluids outside of containment prior to 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months please state whether vital area access (Action Item 11.8.2) necessary to close valves 845, 727 A and 350 and trip the refueling water storage tank (RWST) purification pump is maintained.

Response to Question 7 a) Plant procedure 2-ES-1.3, "Transfer to Cold leg Recirculation", provides instructions for transferring the safety injection system and containment spray system to the recirculation mode. The Procedure requires manually starting one internal recirculation NL-13-115 Docket No. 50-24 7 Page 6of10 pump, and if it cannot be started then manually starting the other internal recirculation pump. If neither internal recirculation pump can be started then the procedure requires establishing cold leg recirculation using RHR pumps which results in sump fluid going outside containment. It should be noted that Emergency Operating Procedures address all potential contingencies to mitigate an accident.

b) The IP2 design is fairly unique in having two internal recirculation pumps as well as two RHR pumps. There is no single active failure that would require using RHR pumps.

Further, 1P2 licensing basis does not postulate a passive failure to occur for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Consequently, recirculation of sump fluid outside containment would only occur at 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for hot leg recirculation. RG 1.183 guidance does not impose postulating a passive failure and consequently ECCS leakage dose is not calculated prior to 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

c) Not Applicable - see response to b) above.

Response To Request For Additional Information Component Performance, NOE. and Testing Branch NL-13-115 Docket No. 50-247 Page 7of10 In the referenced letter it is indicated that valves 845, 727 A and 350 will be part of the lnservice Test Progr~m with a test frequency of two years. Will these valves be classified as manual, active valves and, therefore, be subject to ASME OM Code exercise testing requirements? Will these valves be further classified as Category A and, therefore, be

subject to ASME OM Code leakage testing requirements? (The discussion of post-accident.

dose consequences indicates that these valves could be exposed to sump fluid.)

Response to RAI 1 Valves 845, 727 A and 3.50 will be classified as manual active valves with open and close ASME OM Code exercise stroke requirements on a two yearfrequency. Valves 845 and 727A will be classified as Category A, therefore requiring leak testing every two years.

Valve 350 will not require leak testing. The potential for sump fluid leakage to impact BARS through valve 350 would be leakage through the 2 inch SI 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 in series and tested by Procedure 2-PT-R048, "Leak Test of 842 and 843", and have an acceptance criterion of 0.5 gallons per hour, and accounted for in the radiological analysis.

Question 1 1.0

-INTRODUCTION Response To Request For Additional Information Health Physics and Human Performance Branch NL-13-115 Docket No. 50-247 Page 8of10*

By letter dated April 15, 2013 (ADAMS Accession Number ML13116A007), Entergy Nuclear Northeast (Entergy), licensee for Indian Point Nuclear Generating Unit 2 (IP2), submitted a license amendment request (LAR) to revise Technical Specification (TS) 3.5.4, "Refueling Water Storage Tank (RWST)". The proposed change would revise the TS to allow the non-seismically qualified piping of the temporary Boric Acid Recovery System (BARS) to be connected to, and isolated from, the RWST's seismically qualified piping by manual operation of RWST seismically qualified boundary valves. This would be done under administrative controls and only for limited periods of time. These limited periods are specified as up to 30 days per fuel cycle for filtration for removal of suspended solids from the RWST water. This change will only be applicable until Refueling Outage R22 (Spring 2016) ends. Manual connection of the RWST seismically qualified piping to non-seismically qualified piping shall not be allowed after the end of R22. The Health Physics and Human Performance Branch (AHPB) has done a preliminary review of the LAR regarding the operator performance aspects and finds that the following additional information is required to complete the review.

1. As described in Section 2 of the licensee's submittal, the change requested for TS 3.5.4 is a proposed Note, that states," 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." Later in Section 3, it is stated that, "Prior to refueling outage (RO) 2R20 the RWST was recirculated for a duration of 13 days. After recirculation the total concentration of silica was less than 1.1 ppm. Prior to RO 2R19 the RWST was recirculated for a duration of 11 days. A sample taken after recirculation had total concentration of silica of 1.3 ppm." Based on this statement the NRC staff assumes that clarity was sufficient after, at most, 13 days, and 'at a silica concentration of 1.3 ppm.

a. What concentration of silica/clarity is acceptable for operators to perform their required tasks during shutdown? Why isn't this criterion included in the proposed TS? How will operators know when it is okay to disengage the BARS?

b. If prior to the previous two refueling outages, it only took 11 days and 13 days to achieve acceptable clarity, why is the licensee requesting allowance for up to 30 days?

In order to minimize the time spent in a seismically vulnerable configuration, why wouldn't a duration of 15 days be sufficient?

Response to Question 1

a. The fuel vendor has specified guidelines for implementing zinc addition. For IP2, Chemistry Procedures specify a silica concentration of s 2 ppm to reduce zinc silicate precipitation on fuel surfaces. This is a fuel vendor guidance value, and not a limiting condition for operation. Exceeding this limit would result in fuel exams. Chemistry NL-13-115 Docket No. 50-247 Page 9of10 monitors silica and boron every six hours during the clean up, and is able to predict completion time a day or two ahead of reaching the target value.

b. The 11 days and 13 days in the prior two outages was BARS system operation time.

Time is also required for setup and removal of the BARS skid, which is typically one or two days each. Plus there is a period when the BARS unit is secured but still connected to allow the vendor some time off. The 30 days request provides margin in consideration of any delays or equipment issues that might arise with the vendor skid.

Since the BARS skid is rented, typically for 21 days, it is only used for the amount of time it is needed.

Question 2 I

Does IP2 have a Time-critical Action Program to protect high-risk, time-limited actions from inadvertent change? If yes, is the proposed task sequence included in that program? If no, what controls are used to prevent inadvertent changes to the proposed operator actions or the time available to perform them? Does the licensee's configuration control system have a way to identify Tech-Spec-related actions in procedures?

Response to Question 2 IPEC has a Time-critical Action Program, OAP-115,"0perations Commitments and Policy Details". Specific IP2 actions are listed in Attachment 4, however, the proposed task sequence is currently not included in that program. Licensing Request LR.:LAR-2013-00113 CA#12 has been initiated to update OAP-115 prior to implementing BARS to include an action to isolate BARS in 31 minutes in the event of a seismic occurrence or an accident requiring injection from the RWST. Further, a CAUTION in 2-0SP-10.1.1, "Support Procedure - Safety Injection Accumulators and Refueling Water Storage Tank Operations", specifies the time available to the dedicated operator to isolate the RWST Silica Cleanup System in the event of a failure such that RWST level will be maintained above the Technical Specification limit. Revisions to Procedures require a Process Appliccibility Determination be performed which would evaluate the affect or potential affect of the change.

Question 3 In the general discussion of the ingress/egress paths taken by the operators to accomplish the isolation of seismic from non-seismic systems, the licensee states that a card reader is in the intended path.

a. Does this card reader require a different card than an operator would have for plant access? If yes, will the dedicated operator routinely keep this other card on his person?

If no, where will it be stored?

b. Did the simulation that was performed to ascertain required time vs. available time include accessing the card reader?

c. Is the card reader designed to work under seismic conditions? SBO? How much additional time would be involved if the operator had to deal with a non-operational card reader?

Response to Question 3 NL-13-115 Docket No. 50-247 Page 10of10

a. No. The card reader uses the employee ID card (security badge), which is the same card as an operator would have for plant access. When at work, company policy requires all employees to wear their ID card on the outside clothing, between the neck*

and waist.*

b. Yes. The simulation included accessing the card reader.

c.

No. The security access card reader system is not seismic and will not work under SBO. Operators have keys in their possession to provide manual override in the event of a non-functional card reader and would result in minimal additional time to open the door. As noted in the submittal, a simulation performed by Operations demonstrated substantial margin in the time available to shutdown the BARS and maintain RWST level within the TS value.

Question 4 What method(s) will be used to monitor the continuing effectiveness and safety of the current method of purification of reactor water until the final resolution is implemented in 2016? Will the Corrective Action Program be used to track the status and effectiveness of current process?

Response to Question 4 The continuing effectiveness and safety of the current method of purification of reactor water is monitored by the work control and temporary alteration processes. The Corrective Action Program is used to document and resolve issues that may arise during the campaign.

EN-DC-141, Design Inputs Page 1 of 9 ATTACHMENT9.1 DESIGN INPUT RECORD Sheet 1 of 1 Design Input Revision: 0 I Page 1 oE 9 DESI"" INPU'l' RECORD Document 'rype:

ca1,.., lation Document Number:

TP-c~*.r*-11-nr111H">1 I

}II r,-

Document Revision:

0 P i;:Qbl ~m sum.'t\\arl:

(l\\,tt.;ir;;!:l, adgitioa11.l Sh§§!;§ as r§guiredl The high head safety injection (HHS!) system and the low head injection/residual heat removal (RHR) system are connected to the refueling water storage tank (RWST) through multiple valves.

The potential doses resulting from leakage of-the emergency core cooling system (ECCS) back to the RWST through these valves need to be quantified based on alternate source term (AST) analyses for a large break loss-of-coolant accident (LOCA).

D~sign Qbj§~tive:

{Att2ch 9dditiongl shee!:;s as !:!i!!J11.i;i;:ed)

This analysis wi 11 calculate the Indian Point Unit 2 (IP2) Control Room (CR), off site and the Technical Support Center (TSC) doses resulting from the identified emergency core cooling system (ECCS) back leakage to the IP2 Refueling Water Storage Tank (RWST) during a large break Loss-Of-Coolant Accident (LOCA).

The calculated dose due to the ECCS back leakage to the RWST wi 11 be combined with the calculated dose resulting from releases via the containment leakage and the ECCS recirculation leakage pathways.

Di~~inling B~v1~~;

Contributing Disciplines:

Pr*epared Reviewed Prepared By Reviewed By:

Bv Bv:

0Mechanical 0Electrical Dr & c Ocivil/Structural

Other

!:L.

0Engineering (Nuclear)

,r.E. Chancr Golsn~n'./.

-P?J/ ~ ?rograros

~.,,,...,,,.,.,,

r~

Outside Design-Agency....::....__.

v ODA Responsible Engineer (Print/Sign/Date)

'l'he contributing discipline engineer shall provide his/her na11e beside the appropriate block.

Lead Discipline Fusi ls ~ N~1gl§lar

~-

An.al:t§.it>

z~-t:

l oL:a \\ l ll RE: (Print/Sign)

Mehdi Golshani Date r

Enaineerinq Supervisor:

Ardes;.r T~;,ni ~~

~Ao.-.. ~

\\ ol~1 /II Date EN-DC-141 Rev. 10

EN-DC-141, Design Inputs Page 2 of 9 ECCS Back leakage to 6.S hours*

Reference:

CN-llS..03-8, Rev. o. lndi~n Point Unit 2 {tPP) Upr~te Post-LOCA Calculations."

the RWST -start of leakage sump WaterVolume 374,000 From Page 56 of Reference, CN-CRA-03~55, r'lndian Polnt 2-lOCA Doses for.Stretch Power Upratet>ro~ram/' Revision O, (10/31/03)

gallons Density of RWST water 61.$6 lb/ft$

Density of water at maximum temperature.of 110 *f, from Reference CN*CRA-03~55, "Indian Point 2 -LOCA Ooses Flow Rate ofECCS Ba~k leakage to the RWST -

below the water level Flow Rate ofECCS Back Leakage to the RWST -

above the water level Volume of Water Assoc:iated with ECCS Back leakage to the RWST for Stretch Power Uprate Program," Revish:m o, (10/31/0l)i Page 56 20 & 29 gallons Calculate allowable back leakage to remain below the F~R c~ TEDE per hcmr (gph) dose limit. Also. calculate value which does not result the dose acceptance limit.

Not Applicable The ECCS back leQkage to the RWST above the water level will be considered as part of the ECCS !gakage in the Primarv Auxiliary Building yia the contalnment vent in CN..CRA-03-SS, Revision 0. Note that the RWST releases are bounded by the Primary Auxiliary Building releases since the atmospheric dispersion factors of the PAB releases

{via containment vent) are greater' than those of the RWST releases.

(See inputs for atmospheric dispersion factors (yQ's)J 1,880gaUons

Reference:

IP*CAlC-11-00063, Table 2. The minimum water volume is estimated to be 2,094 (1943 + 151) gatlons between the high head safety injection pump suction and the valve 846 to the RWST 2094 gallons x 0.9 (10% margin)= 1,884.6 sa.llons 1'lt 1,880 gallons Mass ofiodine in sump 26,12lg

Reference:

CN*CRA-11-25, "Indian Polnt3 LOO\\ Doses Including Contribution from Back*leakage to RWST," IP..C:AlC>-11-00080,

{9/23/2011)

Both plants IP2 and IP3 have the same rated thermal Power ancUhe source inventory of the core is almost the same; Therefore the amount of iodine source in the core for both plants tP2 and IP3 should be almost the same.

EN-DC-141 Rev. 10

EN-DC-141, Design Inputs ECCS Back leakage to**

7$ hot,irs the RWST-time delay

of the sump water 60hours reaching the Rwst t:}as~d on 20and*29 gph assumPtion Iodine speci~s in containment s!Jmp water Elemental:

Organic:

Particulate:

0 0

100 Page 3 of 9

Reference:

lP-CAlC.,11.,00063, Table 2. The horizontal section volume of the piping associated with the ECCS back leakage to. the RWST is estimated to be 1,931 gallons b~tween the ~igh, head safety lfijedion pump suction and the valve 846 to the RWST. Sin.ce lhe surnpwatertemperature is higher than t~eRWSTand its associated piping temperatures, and the sump water Is located at lower elevation, the vertical sections of piping are neglected d.ue.to the buoyancy-driven thermal mixing. Therefore, the th;ne:delay of the ECCS back leakage of20 or 29 gallons per hour {gph} to reach the RWSt is conservatively estimated to be 75 or 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months after the start of ECCSexte.rnal re<;lrcufation.

Horizontal Sections of 1A and 18:

429+51+1304 + 78+3 +3 + 40 + 25+30+19 + lQ gallons =

1992 gallons

. Horizontal Sections of 2A and lB:

429+51+1304 + 78 +36 + 283 + 47 + 211 + 17 gallons

= 2462 gallons 1992 gallons x 0.9 (10% margin):: 11792.8 gallons

~ 1, 750 pl!Ons 1,750 gallons/ 20.gph (assume)= 87.S hours

75 hours 1, 750 gallons/ 29 gph (assume)= 60.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

~oohours All iodine is.assumed to have converted to stable foml In the sump water.

EN-DC-141 Rev. 10

EN-DC-141, Design Inputs Page 4of9 Volume of Water Remaining In the RWST after Recirculation switchover 13900 gallons Lowest RWST Water level = 1.49 ft RWST Minimum Temperature RWST Maximum Temperature 40°F 110°F Post-LOCA RWST 114 °f Maximum temperature due to the ECCS Back leakage to the RWST Reference; CR-1P2-2ooi-d4498 Per IP-PRT-09-00014, Rev, 1, page 45 shows the actual lowest RWST water level is 1.74 and page 42 ofthis reference says "If RWST level decreases to less than 1.5.ftthen stop all pumps taking suction from the RWST/' Therefore, using 1.49.is conservative, RWST vofui'rte; H:::.41'-3" Dia.= 40.0' Drawing No: F~P. No. 9321-01~20339*4 Thickness= 0.221't::::o.t)18917' Volume= TI R2 h= 3.14 x (20-0.018917)2x41.25 = 51738.27 ft3::::

387054.0 gal RWST Water Volume per Foot= 387054.0 / 41.25 = 9383 gal/ft Remaining Water Volume= 1.49 ft x 9383 gal/ft

= 13980 gallons

~ 13,900 gallons SR3.5.4.1 Section 3.5.4 "Refueling Water Storage Tank (RWST)" of Indian Point Unit 2, Improved Technical Spei:ifications (ITS).

SR3.S.4.1 Se~ion3.SA "RefuelingWaterStorage Tank (RWST)" of lridian Point Unit 2, Improved Technical Speclflcations (ITS).

Reference:

CN..CRA.;11..:2s,. "Indian Point 3 LOCA Doses including Contribution from Back*leakage to RWST/' IP-CALC*11-00080, *

(9/23/2011)

The post-LOCA maximum RWST temperature was estimated in Appendix B (pages 79 and 80) of CN-CRA-11-25 for JP3. A review of the IP2 containment s1Jmp temperature and the 'stimated ECCS bacl<~leakage rate concluded that 114 °F is still bounding.

EN-DC-141Rev.10

EN-DC-141, Design Inputs Page 5 of 9 Volume of Air in the 386,000 Height of the RWST = 41' - 3" RWST after Recirtutation. gallons Switchover Drawing No:* F.P. No, 9321-01-20339-4 Ma~lmum Boron Concentration of RWST Minimum BOron Concentration of RWST Maxlt:nmn Diurnal Temperature Variation Minimum Sodium Tetraborate Oecahydrate for Poi;t-LOCA pH Contro I RWSTVolume at 41' -3" = 13820 gallons+ (4i'.... 3;1) x 9383 gallons/ft

=

400,874 galtons Remaining Air Volume= 400,874 gallons -13,980 gallons

= 386,894 gallons

z 386,000 gaHol'ls 26QO parts per SR 3.5.4.3 milfion (ppm]

Section 3.S:.4 "Refueling Water Storage Tank (RWST}" of Indian Point Unit 2, Improved Technlcal Specificati()ns (ITS).

2400 parts per SR 3.5.4.3

. million (ppm]

Secti<;m 3.5A "Refuellng Water Storage Tank (RWST)." of Indian Point Unit 2, Improved Technical Specifications (ITS).

40 °F

. A revie\\Y of the four (4) year tndian floint meteorological data snows the maximum diurnal temperafure VariatiO!) dOe5 OOt exceed 40 Cf.

[See page 81 of IP-CALC*11*o0080, Revision 0, (CN*CRA-11-25, Revision. O), "Indian Point 3 LOCA Doses including Contribution from Back~lmlkage te> RWST."]

8,096 pounds SR 3.6.'7.1.b (lbm)

Section 3.6.7 "Recirculation pH Control System" of Indian Point Unit 2, Improved Technical Specifications (ffS).

EN-DC-141Rev.10

EN-DC-141, Design Inputs Atmospherlc Dispersion

.*Factors [X/UJ for the IP2 Control Room (CR) Air Intake Associated with the IP2 RWST Release

[sec/m3J

.()... 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; 2-s hours:

s.;...;24.hours:

1...:4days!

4...:3qda\\is:

Control Room Volume CR Normal Operation flow rates (cfm)

Filtered Makeup:

Altered Recirci.tlatlon:

Unfiltered Makeup:

.Unfiltered inleakage:

5~62E.-()4 3.72E-04 1.3SE.,.Q4 L10E-04 9.02E-OS 1()2,400 ft3 0

()

920 100 Time to sWitch CR HVAC 60 sec.

to emergency operation

.m()de CR HVAC emergency operation flow (dm)

Filtered Makeup:

ijnflltered Makeup:

Unfiltered lnleakage;

1800 0

700 Page 6 of 9 Table 2~1 of IP<ALC*ll-000601 Revision o, "Analysis of IP2 Control RQom and TechnlC:ai Support Center Atmospheric Dispersion Factors due to Releases from thelP2 FSB & RWST." (9/2$/11)

Consistent with analysis in Reference i CN-CRA~o3.;;55, "lndia.n Point 2 - LOO\\' Doses for Stretch Power Uprate Program, 11 Revision O, (10/31/03)

Consistent with analysisJn Reference, CN~CRA-03*55, n1ndia.n Poirit 2 - LOO\\ Doses forStret~h Power Uprate Program/

Revision 0, (10/31/03)

Consistent with analysis In Reference, CN..CRA*03.,55, "Indian Point 2-LOCA Doses for Stretch Power Uprate Program,"

Revision o, (10/31/03)

Consistentwith analysis in Reference, CN..;CRA~03:-5S,*"lndian Point 2 - LOCAOoses for Stretch Power Uprate Program/'

Revision 0, (10/31/03)

EN-DC-141 Rev. 10

EN-DC-141, Design Inputs CR HVAC Filter efficiencies (%)

Elemental iodine:

Organic iodine:

Particulates:

CR Breathing rate (m3/sec)

CR Occupancy Factors 0-2 hours:

1-4 days:

4-30 days:

Offsite Meteorological Dispersion Factors (sec/m3)

EAB 0-2 hours:

LPZ 0-8 hours:

8-24 hours:

1-4 days:

4-30 days:

Offsite breathing rate (m3/sec) 0-8 hours:

8-24 hours:

1-30 days:

Technical Support Center {TSC) Net-free Volume 95 90 99 3.SE-04 1.0 0.6 0.4 7.SE-04 3.SE-04 1.2E-04 4.2E-05 9.3E-06 3.SE-04 1.8E-04 2.3E-04 860.9 m3 Page 7 of 9 Consistent with analysis in Reference, CN-CRA-03-55, "Indian Point 2 - LOCA Doses for Stretch Power Uprate Program,"

Revision 0, (10/31/03)

Ref. Reg. Guide 1.183 and also consistence with analysis in Reference, CN-CRA-03-55, "Indian Point 2 - LOCA Doses for Stretch Power Uprate Program," Revision 0, (10/31/03)

Consistence with analysis in Reference, CN-CRA-03-55, "Indian Point 2 - LOCA Doses for Stretch Power Uprate Program,"

Revision 0, {10/31/03)

Ref. Reg. Guide 1.183 and also consistence with analysis in Reference, "Indian Point 2 - LOCA Doses for Stretch Power Uprate Program," Revision 0, (10/31/03)

Page 11 of NEA-00023, Revision 0, "Unit 2 TSC Personnel doses from RG 1.183/NUREG-1456 Design Basis Loss-of-Coolant Accident."

EN-DC-141Rev.10

EN-DC-141, Design Inputs Page 8 of 9 Atmospheric Dispersion

Factors {y}Ql for the Technh:al Support Center (TSC) Air Intake Associated with the IP2 RWST Rel.ease [sec/m3]

o.....;2 hour$:

2--8 hours:

8-24 h.ours:

1-4 days:.

4-30days:

3.SSE-04

.1;24E*04

S;66E-05 4;77E..05 3.94E..05 Table 2.2 of IP~CALC~11~oooso, Revislon o, '!Analysis of IP2 Contrpl Room and Technical.Support Center Atmospheric Oisperston Factors due to Releases from the lP2 FSB & RWST." (9/28/11)

Technlcal Support 12,870 cfm 8620 cfm + 4250 cfm = 12,870 cfm

<;enter (TSC) Unfiltered Intake flowRate

[Normal Operation)

Technical SupPQrt Center (TSC) Ventilation Mode* (Incident Operation]

Techni~I Support Center (TsC) Filtered Intake Flow Rate

[Incident Operation]

This value is greater than 11,230 cfm [damper flow rate} and 12~500 cfm {air-handling fan flow rate] for conservatism.

A2265861Revision6, "TechniCal SupportCenterHVACFlow Diagram Elev. 72'..0", Elev. 88'-6" (Unit #2).

A226587, Revision 3, "Technical Support Center HVAC Flow Diagram El. 33'*0", 37'-0" & 53'-0" (Unit#2)."

Filtered.

Page 11 of NEA-00023, Revision o, "Unit 2 TSC Personnel doses from pres!>urized RG 1.183/NUREG-1456 Oesisn Basis toss-of~oolant Accident."

intake 3400 standard 3492 to 4268 dm:

cubic feet per minute (scfm) 2-PT*EMOOl, Revision o, "TSC Filtration System."

[conservatively lowered from 3492 dm]

3770 scfm:

Page 11 of NEA-00023, Revision 0, "Unit 2 lSC Personnel doses from RG l.183/NUREG-145Q Design Basis loss~of-Coolant Accident."

EN-DC-141Rev.10

EN-DC-141, Design Inputs Technical Support Center {TSC)

Recirc1.datiori flow Rate

[Both Normal and Incident Operation)

Technical Support Center (TSC) Ventllaticm Mode Change from Normal fo lilddent Operatll.}n Osdm[No Redrculatlon1 6() minutes

. [maximum delay time for conservatism)

Technical Support SOO scfm Center (TSC) Unfiltered lnleakage Ftow Rate (Both Normal and Incident Operation]

TechnieafSupport 3900scfm Center (TSCl Exhaust Flow Rate [lntident Operation]

Technical Support Center (TSC} Filter Efficiencies [Incident*

Operation]

Particulate:

lnorganlts {elemental):

Organics:

Noble gases~

Page 9 of 9 Page.11 of NEA-0002~. Revision 0# "Unit 2 TSC Personnel doses from RG l~i83/NUREG-14S6 Design Basis Loss-of-Coolant Actident/'

The Technical Support Center (TSC) and the Operations Support

eenter (OSCJ will be staffed Within 60minutes, and the osc Radiation Protection Coordinatorwlll request the Control Room to aliSl'.l the TSCventitation system for incident.operation.

IP-EP-210, Revision 9, "Central Control Room.'

1 IP~ep..:220, Revision 10, "Technical Support Center."

IP-EP..:230, Revision 7, "Operations Support Center."

Page 11 of NEA.;00023, Rision 0, "Unit 2 TSC Personnel doses from RG 1.183/NUREG~145~ De$ign Basis loss*of-COolant Accident!'

3400 scftn {filtered lntakej + 500 sdm [unfiltered inleakageJ

=3900scfm Page 11 of NEA-00023, Revtsion 0, "Unit 2 TSC Personnel doses from RG 1.183/NUREG*14SG Design Basis Loss~of*Coolant Acddent~'

1 EN-DC-141 Rev. 10

ATTACHMENT 9.1 Sheet 1of1 D AN0-1 OPNPS Page 1 oflO DESIGN VERIFICATION COVER PAGE DESIGN VERIFICATION COVER PAGE D AN0-2 ovv r8J IP-2 D GGNS D IP-3 ORBS OJAF 0W3 OPLP ONP Document No.: IP-CALC-11-00091 Revision No.: 0 Page 1of10

Title:

AST Analysis of IP2 to Address the Impact of Containment Sump Solution Back-Leakage to the RWST after LOCA OV Method:

r8l Quality Related r8l Design Review VERIFICATION REQUIRED Originator:

0 Augmented Quality Related 0 Alternate Calculation 0 Qualification Testing DISCIPLINE Electrical Mechanical Instrument and Control Civil/Structural Nuclear VERIFICATION COMPLETE AND COMMENTS RESOLVED (DV print, sign, and date Jong E. Chang I Print/Sign/Date After Comments Have Been Resolved EN-OC-134 REV 4

Page 2 oflO ATTACHMENT 9.6 DESIGN VERIFICATION CHECKLIST Sheet 1of3 IDENTIFICATION:

DISCIPLINE:

DocumentTitfe: AST Analysis of IP2 to Address the Impact of Containment Sump OCiviVStructural OElectrical Solution Back-Leakage to the RWST after LOCA Doc. No.: IP-CALC*11-00091 Rev. 0 QA Cat.

Ol&C

-:::.;---~

11/16/u 0Mechanical Jong E. Chang 181Nuclear Verifier:

Print v-si"an

' Date 00ther Manager authorization for supervisor performing Verification.

181 NIA Print Sign Date METHOD OF VERIFICATION:

Design Review t8I Alternate Calculations 0 Qualification Test 0 The following basic questions are addressed as applicable, during the performance of any design verification. [ANSI N45.2.11-1974] [NP QAPD, Part II, Section 3] [NP NQA-1*1994, Part I, BR 3, Supplement 3S*1]

NOTE The reviewer can use the "Comments/Continuation sheer' at the end for entering any comment/resolution along with the appropriate question number. Additional items with new question numbers can also be entered.

1.

Design Inputs-Were the inputs correctly selected and incorporated into the design?

(Design inputs include design bases, plant operational conditions, performance requirements, regulatory requirements and commitments, codes, standards, field data, etc.

All information used as design inputs should have been reviewed and approved by the responsible design organization, as applicable.

All inputs need to be retrievable or excerpts of documents used should be attached.

See site specific design input procedures for guidance in identifying inputs.}

Yes 181 '

No 0 N/A 0

2.

Assumptions-Are assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are assumptions identified for subsequent re-verification when the detailed activities are completed? Are the latest applicable revisions of design documents utilized?

Yes 181 No 0 N/A 0

3..

Quality Assurance - Are the appropriate quality and quality assurance requirements specified?

Yes 181 No 0 N/A 0 EN-DC-134 REV 4

Page 3of10 ATTACHMENT9.6 DESIGN VERIFICATION CHECKLIST Sheet 2 of 3

4.

Codes, Standards and Regulatory Requirements - Are the applicable codes, standards and regulatory requirements, including issue and addenda properly identified and are their requirements for design met?

Yes 181 No 0 NIA 0

5.

Construction and Operating Experience - Have applicable construction and operating experience been considered?

Yes 0 No 0 N/A 181

6.

Interfaces - Have the design interface requirements been satisfied and documented?

Yes 181 No 0 N/A 0

7.

Methods - Was an appropriate design or analytical (for calculations) method used? *

. Yes 181 No 0 NIA 0

8.

Design Outputs - Is the output reasonable compared to the inputs?

Yes 181 No 0 N/A 0

9.

Parts, Equipment and Processes -Are the specified parts, equipment, and processes suitable for the required application?

Yes 0 No 0 N/A 181

10.

Materials Compatibility -Are the specified materials compatible with each other and the design environmental conditions to which the material will be exposed?

Yes 0 No 0 NIA 181

11.

Maintenance requirements - Have adequate maintenance features and requirements been specified?

Yes 0 No 0 N/A 181

12.

Accessibility for Maintena.nce -Are accessibility and other design provisions adequate for performance of needed maintenance and repair?

Yes 0 No 0 N/A 181

13.

Accessibility for In-service Inspection - Has adequate accessibility been provided to perform the in-service inspection expected to be required during the plant life?

Yes 0 No 0 NIA 181

14.

Radiation Exposure - Has the design properly considered radiation exposure to the public and plant personnel?

Yes 181 No D N/A 0

15.

Acceptance Criteria-Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisfactorily accomplished?

Yes 181 No D NIA 0

16.

Test Requirements - Have adequate pre-operational and subsequent periodic test requirements been appropriately specified?

Yes 0 No D N/A 181 EN-DC-134 REV 4

Page 4of10 ATIACHMENT9.6 DESIGN VERIFICATION CHECKLIST Sheet 3 of 3

17.

Handling, Storage, Cleaning and Shipping -Are adequate handling, storage, cleaning and shipping requirements specified?

Yes 0 No 0 NIA 181

18.

Identification Requirements - Are adequate identification requirements specified?

Yes 0 No 0 NIA 181

19.

Records and Documentation -Are requirements for record preparation, review, approval, retention, etc., adequately specified? Are all documents prepared in a clear legible manner suitable for microfilming and/or other documentation storage method? Have all impacted documents been identified for update as necessary?

Yes 181 No 0 NIA 0

20.

Software Quality Assurance-ENN sites: For a calculation that utilized software applications (e.g., GOTHIC, SYMCORD), was it properly verified and validated in accordance with EN-IT-104 or previous site SQA Program?

ENS sites: This is an EN-IT~104 task. However, per ENS-DC-126, for exempt software, was it verified in the calculation?

Yes 181 No 0 NIA 0

21.

Has adverse impact on peripheral components and systems, outside the boundary of the document being verified, been considered?

Yes 0 No 0 NIA 181 EN-DC-134 REV 4

(

Page 5of10

.ATTACHMENT 9.7 DESIGN VERIFICATION COMMENT SHEET Sheet 1of1 Comments I Continuation Sheet Question Comments Resolution Initial/Date 1

Various editorial comments were N/A

j~t!_

identified and addressed. No response I"~ u/1t/11 required.

2

[Section 6.2] The potential ECCS NIA back-leakage flow to the RWST via

/IJ6- :r-zL Valve 846 is below the RWST water level as identified in Attachment C, U//b( If Design Inputs (page 2 of 9). Hence, the flow to "air' is an artificial flow path to match the iodine partition factor between the RWST water and the RWST air. This is not the potential ECCS leakage to the RWST via MOV-842 and MOV-843, which is above the RWST water level. No response required.

3

[Section 6.3] The equivalent mole of NIA M(-. fl{_

sodium hydroxide (NaOH) was used to determine the delivered sump water.

(( /!6 /11 Then the titration curve of boric acid/trisodium phosphate (TSP) was used to estimate the RWST pH.

However, the actual sump solution is based on sodium tetraborate (STB),

which is also a weak base. Therefore, it is not obvious that the titration curve using TSP is always conservative to estimate the RWST pH for the elemental iodine fraction.

The following post-LOCA sump pH values are found based on TSP and STB:

12,000 lbm of TSP (trisodium phosphate dodecahydrate, TSP-1 OH20) in the post-accident IP2 containment sump at 2000 ppm boron resulted in pH of 7.61 (page 10 of CN-CRA-96-005, Revision 2). If the mass is adjusted to 10,000 lbm otTSP, the resulting pH is approximately 7.53.

10,000 lbm of ST.B (sodium EN-DC-134 REV 4

Page 6of10 Question Comments Resolution Initial/bate I

tetraborate decahydrate, STB-12H20) in the post-accident IP2 containment sump at 2000 ppm born resulted in pH of 7.4 (Figure 3 of IP-CALC -

00129, Revision 2).

The molar mass of TSP-1 OH20 is 380.1234 g/mol and STB-12H20 is 381.38 g/mol so they are very comparable in weight. Thanks to the TSP titration curve, the estimated RWST pH could be higher as much as pH = 0.13, which is non-conservative.

While determining the elemental iodine fraction in page 31 of the calculation, the elemental iodine fraction in the RWST was selected based on pH of 6.04 instead of 5.2, which gives a margin of pH = 0.84.

Therefore, although the TSP titration curve results in slightly non-conservative RWST pH, the elemental I

iodine fraction was chosen such that the inputs to RADTRAD are still rconservative.

No response required.

4

[Design Inputs] The maximum RWST N/A M(j-

-::f ~L temperature was reviewed not just for 11/!6/tr the final temperature but for the whole accident duration, i.e., 30 days. As shown in the following Attachment 1, the maximum RWST temperature maintains below 114 °F at 20 gph of the back-leakage flow rate. No

~ response reQuired.

EN-DC-134 REV 4

Page 7of10. Maximum RWST Water Temperature due to the Sump Water Back-Leakage INPUTS t_initial =

start of ECCS leakage to RWST t_final =

6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 23,400 seconds accident duration 30 days 2,592,000 seconds V _pipe=

piping volume 1880 gallons V _rwst =

RWST water volume 13900 gallons T_rwst =

RWST water temperature 110 deg F Q_leakage

=

ECCS leakage to RWST 20 gallons per hour Maximum ECCS Temperature

Minimum ECCS w/ NUREG-1465 Time ECCS Temperature

[sec]

[deg F]

23,199 196.31 25,599 191.26 26,799 188.97 29,199 184.86 31,599 191.28 36,399 175.45 41,599 170.68 61,199 160.83 80,799 156.51 85,599 155.78

Section 2.0 of CN-LIS-30-8, Revision 0

Section 3.1 of CN-CRA-03-55, Revision 0

Regulatory Guide 1.183

Design Input

RHR Suction Line

Design Input

SR 3.5.4.1 of IP3 Improved Technical Specifications

Design Input

pages 61-63, CN-CRA-03-12, Revision 0

Sump Temperature EN-DC-134 REV 4

90,399 155.11 99,999 153.84 101,999 151.64 104,999 148.76 106,999 147.11 109,999 144.94 114,999 142.05 119,999 139.84 128,999 137.05 138,999 135.06 158,999 132.81 199,999 130.53 201,999 129.66 206,999 127.71 216,999 125.11 236,999 122.59 275,999 120.89 314,999 120.09 353,999 119.46 401,999 118.54 411,999 117.4 431,999 116.32 470,999 115.67 548,999 115.21 626,999 114.87 782,999 114.27 1,008,999 111.09 1,094,999 107.04 1,251,999 106.82 1,854,999 106.62 3,750,999 106.09 CALCULATION ECCS Back Leakage to RWST

[sec]

[deg F]

23400 196.31 25599 191.26 26799 188.97 29199 184.86 31599 191.28 36399 175.45 41599 170.68

[gallon-deg F]

2398.3 1275.1 2519.6 2464.8 5100.8 5068.6 18585.2

[gallon-deg F]

2398.3 3673.3 6192.9 8657.7 13758.S 18827.1 37412.2 Page 8of10

[gallons]

12.2 18.9 32.2.

45.6.

[deg F]

£~1fit EN-DC-134 REV 4

/

Page 9of10 61199 160.83 17512.6 54924.8

. 318.9

~'i 80799 156.51 4173.6 59098.4 345.6 85599 155.78 4154.1 63252.6 372.2 90399 155.11 8272.5 71525.1 425.6 99999 153.84 1709.3 73234.4 436.7 101999 151.64 2527.3 75761.8 453.3

~r,.'.

104999 148.76 1652.9 77414.7 106999 147.11 2451.8 79866.5 481.1.;/

109999 144.94 4026.1 83892.6 508.9 114999 142.05 3945.8 87838.4 536.7 119999 139.84 6992.0 94830.4 586.7 128999 137.05 7613.9 102444.3 642.2 138999 135.06 15006.7 117451.0 753.3 158999 132.81 30251.2 147702.2 981.1 199999 130.53 1450.3 149152.5 992.2 201999 129.66 3601.7 152754.2 1020.0 206999 127.71 7095.0 159849.2.

1075.6 216999 125.11 13901.1 173750.3 1186.7 236999 122.59 26561.2 200311.4 1403.3 275999 120.89 26192.8 226504.3 1620.0 314999 120.09 26019.5 252523.8 1836.7 353999 119.46 31856.0 284379.8 2103.3 401999 118.54 6585.6 290965.3 2158.9 411999 117.4 13044.4 304009.8 2270.0 431999 116.32 25202.7 329212.4 2486.7 470999 115.67 50123.7 379336.1 2920.0 548999 115.21 49924.3 429260.4 3353.3 626999 114.87 99554.0 528814.4 4220.0 782999 114.27 143472.3 672286.8 5475.6 1008999 111.09

)

53076.3 725363.1 5953.3 1094999 107.04 93362.7 818725.8 6825.6 1251999 106.82 357847.0 1176572.8 10175.6 1854999 106.62 436550.3 1613123.0 14270.0 2592000 106.09 V_total =

total water volume Back Leakage 14270 gallons 1613123 gallon-deg F Piping 1880 gallons 206800 gallon-deg F RWST 13900 gallons 1529000 gallon-deg F Total=

30050 gallons 3348923 gallon-deg F EN-DC-134 REV 4

Page 10of10 T _final =

final RWST temperature 111.445 deg F 112 degf The final RWST temperature is conservatively increased to 114 °F.

EN-DC-134 REV 4

v t e Letter Sequence Other
TAC:MF1440, (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions 7 August 2013 Answers 4 September 2013 Results Approval Other: ML14002A431, ML17104A041, ML17104A042, NL-13-015, Proposed License Amendment Regarding Connection of Non Seismic Boric Acid Recovery System to the Refueling Water Storage Tank, NL-17-035, Proposed License Amendment Regarding Connection of Non-Seismic Boric Acid Recovery System to the Refueling Water Storage Tank
MONTHYEAR2013-05-30 [Table View]

NL-17-035, Proposed License Amendment Regarding Connection of Non-Seismic Boric Acid Recovery System to the Refueling Water Storage Tank

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

Enclosure:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

5.0 REGULATORY ANALYSIS

7.0 REFERENCES

SUBJECT:

REFERENCES:

Dear Sir or Madam:

Attachment:

Enclosure:

Reference:

Reference:

Reference:

Reference:

Reference:

Title:

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