Text

License Amendment Request Pursuant to 10 CFR 50.90: Maximum Extended Load Line Limit Analysis Plus - Revision 1
	ML14169A034
Person / Time
Site:	Nine Mile Point
Issue date:	06/13/2014
From:	Costanzo C; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TAC ME3145, TAC MF2462
	Download: ML14169A034 (46)
	v • d • e

AChris Costanzo AM P--i Site Vice President - Nine Mile Point ExeonG P.O. Box 63 Lycoming, NY 13093 315-349-5200 Office www.exeloncorp.com Christopher.costanzo@exeloncorp.com June 13, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69 Docket No. 50-410

Subject:

License Amendment Request Pursuant to 10 CFR 50.90: Maximum Extended Load Line Limit Analysis Plus - Revision 1

References:

(1) Letter from P. Swift (NMPNS) to Document Control Desk (USNRC),

License Amendment Request Pursuant to 10 CFR 50.90: Maximum Extended Load Line Limit Analysis Plus, dated November 1, 2013 (ADAMS Accession No. ML13316B1307)

(2) Letter from B. Vaidya (USNRC) to C. Costanzo (NMPNS), Nine Mile Point Nuclear Station, Unit No. 2 - Issuance of Amendment, Re:

License Amendment Request Pursuant to 10 CFR 50.90: Standby Liquid Control System - Increase in Isotopic Enrichment of Boron-10 (TAC No. MF2462), dated March 14, 2014 (ADAMS Accession No. ML14036A005)

(3) Letter from J. Stanley (NMPNS) to Document Control Desk (USNRC), License Amendment Request Pursuant to 10 CFR 50.90:

Maximum Extended Load Line Limit Analysis Plus - Response to RAI STSB-1 and RAI STSB-2, dated May 14, 2014 (ADAMS Accession No. ML14139A146)

(4) Letter from T. Beltz (NRC) to K. Fili (Monticello Nuclear Generating Plant), Moniticello Nuclear Generating Plant - Issuance of Amendment No. 180 to Renewed Facility Operating License regarding Maximum Extended Load Line Limit Anaysis Plus (TAC No. ME3145), dated March 28, 2014 (ADAMS Accession No. ML14035A248)

Nine Mile Point Nuclear Station, LLC (NMPNS) hereby transmits Revision 1 to the Nine Mile Point Unit 2 (NMP2) License Amendment Request originally submitted on November 1, 2013 (Reference 1). The request to amend the NMP2 Renewed Facility Operating License No. NPF-69 included a proposed expansion of the operating boundary to allow operation in the Maximum Extended Load Line Limit Analysis Plus (MELLLA

U. S. Nuclear Regulatory Commission June 13, 2014 Page 2 Plus) domain and the use of the General Electric Hitachi Nuclear Energy (GEH) analysis code TRACG04.

Revision 1 of the Enclosure is modified using track changes with all changes in color.

Revision 1 of the Enclosure modifies the original request to reflect:

1) The USNRC's issuance of Amendment No. 143 to NMP2 Renewed Facility Operating License No. NPF-69 regarding an increase in the isotopic enrichment of Boron-10 in the Standby Liquid Control System (Reference 2);

2) Changes to the original LAR made in a response to an USNRC Staff Request for Additional Information (Reference 3);

3) A correction to the No Significant Hazards Consideration regarding the discussion of the change to the long term stability solution; and

4) The USNRC's issuance of Amendment No. 180 to Monticello Nuclear Generating Plant's Renewed Facility Operating License No. DPR-22.

Revisions to the Technical Specification (TS) and TS Bases were previously submitted in Reference (3). The changes in the TS and TS Bases have been incorporated into revision 1 of the Enclosure to provide a marked up copy of the Enclosure encompassing all changes to date. Attachment 1 to the Enclosure includes a revision to TS page 3.1.7-3 which replaces the corresponding page previously submitted in Reference (1). TS page 3.1.7-4 and associated TS Insert 1 - Figure 3.1.7-1 were implemented with approval of NMP2 Amendment 143, and are to be removed from the original Attachment 1 submitted in Reference (1). The remaining pages in Attachment 1 to the Enclosure, submitted on November 1, 2013 (Reference 1) and modified by the submittal on May 14, 2014 (Reference 3), are not changed with this submittal. Attachments 2 through 11 of the Enclosure to the original request submitted on November 1, 2013 (Reference 1) are not revised or reissued.

This submittal revises the No Significant Hazards Determination analysis provided by NMPNS in Reference (1). Pursuant to 10 CFR 50.91 (b)(1), NMPNS has provided a copy of this supplemental information to the appropriate state representative.

This letter contains no new regulatory commitments.

Should you have any questions regarding the information in this submittal, please contact Everett (Chip) Perkins, Director Licensing, at (315) 349-5219.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 1 3 th day of June, 2014.

Sincerely, Christopher R. Costanzo

U. S. Nuclear Regulatory Commission June 13, 2014 Page 3 CRC/KJK

Enclosure:

Revision 1 - Evaluation of the Proposed Changes cc: Regional Administrator, Region I, USNRC Project Manager, USNRC Resident Inspector, USNRC A. L. Peterson, NYSERDA

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGES Nine Mile Point Nuclear Station, LLC June 13, 2014

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Background 2.2 Proposed Changes to the Nine Mile Point Unit 2 Technical Specifications 2.3 Modification Summary

3.0 TECHNICAL EVALUATION

3.1 MELLLA+

3.2 DSS-CD 3.3 Standby Liquid Control System Bororn 10 Enrichment 3.4 Safety Limit Minimum Critical Power Ratio 3.5 NMP2 TS Changes 3.6 TSTF-493 3.7 Topics Discussed During NRC Pre-Meetings

4.0 REGULATORY EVALUATION

4.1 Evaluation of NMP2 License Amendment Requests to Establish that They Are Not Linked 4.2 Applicable Regulatory Requirements/Criteria 4.3 Precedent 4.4 Significant Hazards Consideration 4.5 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

i of ii

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE TABLE OF CONTENTS ATTACHMENTS

1. Nine Mile Point Unit 2 Proposed Changes to Technical Specifications (Mark-ups)

Note: Attachments 2 through II of the Enclosure to the Nine Mile Point Unit 2 License Amendment Request dated November 1, 2013 are not revised or reissued in this revision.

2. Nine Mile Point Unit 2 Changes to Bases for Technical Specifications (Mark-ups)

3. List of Regulatory Commitments

4. MELLLA+ Risk Evaluation

5. Nine Mile Point Unit 2 Power/Flow Operating Map for Current Cycle

6. General Electric - Hitachi Affidavit Justifying Withholding Proprietary Information in NEDC-33576P

7. Global Nuclear Fuel Affidavit Justifying Withholding Proprietary Information in GNF-0000-0156-7490-RO-P

8. NEDC-33576NP, Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load Line Limit Analysis Plus (Non-proprietary)

9. Global Nuclear Fuel Report GNF-0000-0 156-7490-RO-NP, "GNF Additional Information Regarding the Requested Change to the Technical Specification SLMCPR," dated August 26, 2013 (Non-proprietary)

10. NEDC-33576P, Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load Line Limit Analysis Plus (Proprietary)

11. Global Nuclear Fuel Report GNF-0000-0 156-7490-RO-P, "GNF Additional Information Regarding the Requested Change to the Technical Specification SLMCPR," dated August 26, 2013 (Proprietary) ii of ii

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Renewed Operating License (OL) NPF-69 for Nine Mile Point Unit 2 (NMP2). The proposed amendment includes supporting changes to the NMP2 Technical Specifications (TSs) necessary to: 1) implement the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) expanded operating domain; 2) change the stability solution to Detect and Suppress Solution - Confirmation Density (DSS-CD); 3) use the TRACG04 analysis code; and 4) incr.ease the isotpic en.rihment of boro.n 10 in the sodium pentabor.. e so!uti.n utilized in the Standby Liquid Control System (SLS); and 5) increase the Safety Limit Minimum Critical Power Ratio (SLMCPR) for two recirculation loops in operation.

The following is a list of the proposed changes to the NMP2 TSs:

Revise Safety Limit (SL) 2.1.1.2 by increasing the SLMCPR for two recirculation loops in operation from > 1.07 to > 1.09

" Revise the acceptance criterion in TS 3.1.7, "Standby Liquid Control (SLC) System,"

Surveillance Requirement (SR) 3.1.7.7 by increasing the discharge pressure from

> 1,327 pounds per square inch gauge (psig) to > 1,335 psig

" Revise the aeceptance cr-iter-ion int TS SR 3.1.7.10 by inremasing the sodium pentaborate boron 10 enr-iehment Fcquir-ement from -ý25 atom percent to ý! 92 atom percent, and make-a Corespending e hange in TiS Figure 3.1.7 1, "Sodium Pentabortec Solutieo Volumne/Conentration" Requiements"

" Revise TS Figure 3.1.7 I to account for the deer-ease in the miniftmum volume of the SLS t"n gallons and 1,288 gallons at sodium pentabConte from 4,558.6 ionoentations of 13.6oTand 14.49%, respectively, to 1,600 gallons and 1,530 gallons at sodium pentaborete concentrations of 13.6%4 and 14.41%, r-espectively

Change the Required Actions for Condition F of TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"

Change Condition G of TS 3.3. 1.1

Add new Conditions J and K to TS 3.3.1.1

Correct an editorial error in Note 3 to TS SR 3.3.1.1.13 (i.e., "ORRM" is changed to "OPRM")

Eliminate TS SR 3.3.1.1.16 and references to it in TS Table 3.3.1.1-1, "Reactor Protection System Instrumentation"

Change the allowable value (AV) for TS Table 3.3.1.1-1, Function 2.b, Average Power Range Monitor (APRM) - Flow Biased Simulated Thermal Power (STP) - Upscale from "5 0.55W+60.5% [Rated Thermal Power] RTP and 5 115.5% RTP" to

"< 0.61W + 63.4% RTP and < 115.5% RTP"

Add a new note to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased Simulated Thermal Power - Upscale scram setpoint to be reset to the values defined by the Core Operating Limits Report (COLR) to implement the Automated Backup Stability Protection (BSP) Scram Region in accordance with Required Action F-,2-4F.2 of TS 3.3.1.1 I of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE

" Add a new note to TS Table 3.3.1.1-1, Function 2.e, Oscillation Power Range Monitor (OPRM) - Upscale to denote that following implementation of DSS-CD, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered operable and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region

Change the mode of applicability for TS Table 3.3.1.1-1, Function 2.e, OPRM-Upscale from Mode I to > 18% RTP

" Change the allowable value for TS Table 3.3.1.1-1, Function 2.e from "As specified in the COLR" to "NA"

,Add a pr-.hibiti'n t. TS Limiting Condition for Operation (LCO) 3.4.1, "Recirculation Loops Operating," is modified to prohibittimhat-pfhbits operation in the Maximum Extended Load Line Limit Analysis (MELLLA) domain or MELLLA+ expanded operating domain as defined in the COLR when in operation with a single recirculation loop

Add Required Action B.2 to TS 3.4.1 to identify that intentional operation in the MELLLA domain or MELLLA+ domain as defined in the COLR is prohibited when a recirculation loop is declared "not in operation" due to a recirculation loop flow mismatch not within limits

" Revise TS 5.6.5.a.4 to replace "Reactor Protection System Instrumentation Setpoint for the OPRM - Upscale Function Allowable Value for Specification 3.3.1.1" with "The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Simulated Thermal Power - High setpoints used in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1"

" Add TS 5.6.8, "OPRM Report," to define the contents of the report required by new Required Action F-.24F.3 of TS 3.3.1.1 Nuclear Regulatory Commission (NRC) approval of the requested operating domain expansion will allow NMP2 to implement operational changes that will increase operational flexibility for power maneuvering, compensate for fuel depletion, and maintain efficient power distribution in the reactor core without the need for more frequent rod pattern changes. MELLLA+ supports operation of NMP2 at Current Licensed Thermal Power (CLTP) of 3,988 Megawatts - Thermal (MWth) with core flow as low as 85% of rated core flow. By operating in the MELLLA+ domain, a significantly lower number of control rod movements will be required than in the present operating domain. This represents a significant improvement in operating flexibility. It also provides safer operation, because reducing the number of control rod manipulations:

(a) minimizes the likelihood of fuel failures and (b) reduces the likelihood of accidents initiated by reactor maneuvers required to achieve an operating condition where control rods can be withdrawn.

Attachments 8 and 10 provide the non-proprietary and proprietary versions of the MELLLA+

Safety Analysis Report (MELLLA+ SAR), respectively. The MELLLA+ SAR follows the guidelines contained in GE-Hitachi Nuclear Energy Americas (GEH) Licensing Topical Report (LTR) NEDC-33006P-A, Revision 3, "Maximum Extended Load Line Limit Analysis Plus" (MELLLA+ LTR) (Reference 1). The MELLLA+ SAR provides the technical bases for this request and contains an integrated summary of the results of the underlying safety analyses and evaluations performed specifically for the NMP2 expanded operating domain.

The MELLLA+ SAR also provides the analyses to change the NMP2 stability solution from Option III to DSS-CD and use the GEH analysis code TRACG04. DSS-CD as required by the MELLLA+ LTR Safety Evaluation Report. DSS-CD is being implemented using the guidelines 2 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE contained in GEH LTR NEDC-33075P-A, Revision 7, "General Electric Boiling Water Reactor Detect and Suppress Solution - Confirmation Density," (Reference 2). The use of TRACG04 is being implemented using the guidelines contained in GEH LTR "DSS-CD TRACG Application,"

NEDE-33147P-A, Revision 4, August 2013 (Reference 3).

The proposed change to the SLMCPR value for two recirculation loops in operation is based on an analysis performed by Global Nuclear Fuel (GNF) for NMP2 during Cycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO, "GNF Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operation value of SLMCPR from > 1.07 to >_ 1.09, and maintaining the single recirculation loop in operation value of SLMCPR at > 1.09. These values are based on NRC approved methods and procedures.

Attachments 9 and 11 of this Enclosure provide non-proprietary and proprietary versions of the GNF report, respectively.

Attachments 10 and 11 of this Enclosure contain information considered to be proprietary as defined by 10 CFR 2.390. GEH and GNF, as the owners of the proprietary information in Attachments 10 and 11, respectively, have executed the affidavits provided in Attachments 6 and 7 to this Enclosure detailing the reasons for withholding the proprietary information.

Attachment 3 delineates the regulatory commitments associated with the proposed change.

2.0 DETAILED DESCRIPTION 2.1 Background 2.1.1 MELLLA+

Operation of Boiling Water Reactors (BWRs) requires that reactivity balance be maintained to accommodate fuel burn-up. BWR operators have two options to maintain this reactivity balance:

(a) control rod movements or (b) core flow adjustments. Because of the strong void reactivity feedback and its distributed effect through the core, flow adjustments are the preferred reactivity control method. Operation at low-flow conditions at rated power level also increases the fuel capacity factor through spectral shift and the increased flow region compensates for reactivity reduction due to fuel depletion during the operating cycle.

At NMP2, an Extended Power Uprate (EPU) was implemented by extending the MELLLA operating domain up to the EPU power level (3,988 MWth). The extension of the MELLLA line to EPU power levels reduces the available core flow window. In addition, the increased core pressure drop with EPU limits the recirculation flow capability. Consequently, EPU plants generally operate with a reduced core flow window and compensate for reactivity loss with control rod movement. Operation in the MELLLA+ expanded operating domain will provide a larger core flow window for NMP2.

In June 2009, the NRC approved the use of the MELLLA+ LTR (NEDO-33006P-A)

(Reference 1) as a basis for MELLLA+ operating domain expansion license amendment requests, subject to limitations specified in the MELLLA+ LTR and in the associated NRC safety evaluation. The NMP2 request complies with the specified limitations and conditions as discussed in Appendix B of Attachments 8 and 10.

3 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE In January 2008, the NRC approved the use of the DSS-CD LTR (NEDC-33075P) as a basis for implementing DSS-CD as a stability solution to replace the Option III solution in license amendment requests, subject to limitations specified in the DSS-CD LTR and in the associated NRC safety evaluation. The NMP2 request complies with Revision 7 of NEDC-33075P (Reference 2), including the specified limitations and conditions as discussed in Appendix C of Attachments 8 and 10.

The TRACG code for use in DSS-CD applications (NEDE-33147P-A) was approved by NRC in November 2007. The NMP2 request complies with Revision 4 of NEDE-33147P-A (Reference 3).

In addition, the NRC approved the Applicability of GE Methods to Expanded Operating Domain Licensing Topical Report (NEDC-33173P-A) which imposes limitations and requirements for the use of GEH Methods in expanded operating domains including power uprates and MELLLA+

domains. The NMP2 request complies with Revision 4 of NEDC-33173P-A (Reference 4),

including the specified limitations and conditions as discussed in Appendix A of Attachments 8 and 10.

Detailed evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, and design basis accidents were performed and are provided in Attachments 8 and 10. These evaluations demonstrate that NMP2 can safely operate in the MELLLA+ expanded operating domain with DSS-CD as the thermal hydraulic stability solution.

2.1.2 Standby Liquid Control System is.topi. Enrichment of Boron 1 NMPNS proposes to inr-easc the isotopi. Enr'ichm*cnt of beroen 10 in the sodium pentaboate solution used to prcparc the nouttron absortber- solution in the Standby Liquid Control Systom (SLS) to ý!92 atom percoent. The proposed berefn 10 enrielifitnt value allows the minimfum net solution volume storoed in the Slug storage tank to be deer-eased to 1,530 gallons EAt 14.4% sodium pentaberate eoncentraion and 1,600 gallons at 13.6% sodium pentaborae concontration. I additiei-,NMPNS proposes to increase the acceptance criterion for the SLS pump discharge pressure from 1,327 psig toŽ:- 1,335 psig.

2.1.3 Safety Limit Minimum Critical Power Ratio NMPNS proposes to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops in operation from > 1.07 to Ž! 1.09. The proposed change to the SLMCPR value for two recirculation loops in operation is based on an analysis performed by GNF for NMP2 during Cycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO-P, "GNF Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operation value of SLMCPR from Ž! 1.07 to Ž! 1.09, and maintaining the single recirculation loop in operation value of SLMCPR at Žý 1.09. These values are based on NRC approved methods and procedures. Attachments 9 and I11 of this Enclosure provide non-proprietary and proprietary versions of the GNF report, respectively.

2.2 Proposed Changes to the Nine Mile Point Unit 2 Technical Specifications NMP2 TS changes are required to allow operation in the expanded MELLLA+ operating domain, use of DSS-CD, inerease the isotopic enrzeinhmcnt of boroen 10 in the sodium pentaborate solution 4 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE used to prepare the ne.tr.n abs.r.b.r. s.lution in the SLS, and increase the SLMCPR for two recirculation loops in operation. Attachment 1 of this Enclosure provides a mark-up of the NMP2 TS showing the proposed changes. Attachment 2 of this Enclosure provides a mark-up of the NMP2 TS Bases to show the corresponding changes to the TS Bases. Attachment 2 is provided for information only. A description of each TS change is provided below.

Safety Limit 2.1.1.2, Safety Limit Minimum Critical Power Ratio SL 2.1.1.2 is revised to increase the SLMCPR for two recirculation loops in operation from

> 1.07 to Ž 1.09.

TS 3.1.7, Standby Liquid Control (SLC) System TS SR 3.1.7.7 is revised to increase the acceptance criterion for the Standby Liquid Control System (SLS) pump discharge pressure from > 1,327 psig to > 1,335 psig.

T-S ISR 3.167. 10 is r-evised to iner-ease the boron 10 cnrichmcnt requirement of sodium pcntabrtef from ? 25 eatm perceent to ? 92 atom perceent. in addition TS Figure 3.1.7 1 is updated to r-eoc the ineroease in the boront 10 cnriehment roquirement.

TIS Figure 3.1.7 1, "Sodium Pentab-ratm Solution VelumC*!C-nc... ati-n Requir.ements," is revised to aecount for the change in the net volufe in the SL9 tank that arise fro*m the cnr-iehment iner-ease. The miniftmum volumne is changed from 4,558.6 gallonts and 4,288 gallons a sodium pcntaber-ate coneentrationis of 13.6% and 14.4%, respeetively, to 1,600 gaillons and 1,530 gallonis at a sodium pentaberate coneentration of 13.6% and 14.44, r-espeetively.

TS 3.3.1.1, Reactor Protection System (RPS) Instrumentation Required Actions F. 1 and F.2 of TS 3.3.1.1 and their associated Completion Times are replaced with the following new Required Actions and Completion Times.

REQUIRED ACTION COMPLETION TIME F. I Initiate Action to implement the Manual Immediately BSP Regions defined in the COLR.

AND F-.2-4-F._2 Implement the Automated BSP 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Scram Region using the modified APRM Simulated Thermal Power -

High scram setpoints defined in the COLR.

AND 90 dayslmmediately R.-,F.__3 Initiate action in accordance with Specification 5.6.8.

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ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE Condition G is modified to no longer apply in the event a Required Action and associated Completion Time of Condition F is not met.

New Condition J (see below) is added to address the action to take in the event a Required Action and associated Completion Time of Condition F is not met.

New Condition K (see below) is added to address the action to take in the event a Required Action and associated Completion Time of Condition J is not met.

CONDITION REQUIRED ACTION COMPLETION TIME J. Required Action and J. 1 Initiate action to Immediately associated Completion Time implement the Manual of Condition F not met. BSP regions defined in the COLR.

AND J.2 Reduce operation to below 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the BSP Boundary defined in the COLR.

AND J.3 --------- NOTE-------

LCO 3.0.4 is not applicable Restore required channel to 120 days operable.

K. Required Action and K.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion Time POWER to less than 18%

of Condition J not met. RTP "ORRM" is changed to "OPRM" in Note 3 to TS SR 3.3.1.1.13.

TS SR 3.3.1.1.16 is eliminated, and references to it in TS Table 3.3.1.1-1 are eliminated.

TS Table 3.3.1.1-1, Function 2.b, Flow Biased Simulated Thermal Power - Upscale, contains both a flow-biased AV (_ 0.55W + 60.5% RTP) and a fixed AV at 115.5% RTP. The flow-biased AV will be changed to (< 0.6 1W + 63.4% RTP).

A new note is added to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased Simulated Thermal Power - Upscale scram setpoint to be reset to the values defined by the COLR to implement the Automated Backup Stability Protection (BSP) Scram Region in accordance with Required Action F.2. F.2 of TS 3.3.1.1.

A new note is added to TS Table 3.3.1.1-1, Function 2.e, OPRM - Upscale, to denote that following DSS-CD implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes 6 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE completely through the DSS-CD Armed Region. However, DSS-CD is considered operable and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region.

The mode of applicability for TS Table 3.3.1.1-1, Function 2.e, OPRM-Upscale is changed from Mode I to > 18% RTP.

In addition, the allowable value for Function 2.e is changed from "As specified in the COLR" to "4NA."

TS 3.4.1, Recirculation Loops, Operating LCO 3.4.1 is modified to include an additional provision that will prohibit intentional operation in the MELLLA domain or the MELLLA+ domain as defined in the COLR when only a single recirculation loop is in operation. It will state:

"... One recirculation loop shall be in operation provided the plant is not operating in the MELLLA or MELLLA+ domain defined in the COLR and provided the following limits are applied when the associated LCO is applicable:..."

A new Required Action B.2 is added to prohibit intentional operation in the MELLLA domain or the MELLLA+ domain defined in the COLR in the event a recirculation loop is declared to be "not in operation" due to a recirculation loop flow mismatch. The Completion Time for this new Required Action is 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

TS 5.6.5, Core Operating Limits Report (COLR)

TS 5.6.5.a is modified by replacing the reference to "Reactor Protection System Instrumentation Setpoint for the OPRM - Upscale Function Allowable Value for Specification 3.3.1.1" with a reference to "The Manual Backup Stability Protection (BSP) Scram Region (Region 1), the Manual BSP Controlled Entry Region (Region II), the modified APRM Simulated Thermal Power - High setpoints used in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1 ."

TS. 5.6.8, OPRM Report The following new report requirement is added as TS 5.6.8, "OPRM Report:"

"When a report is required by Required Action F-,2-F.3 of TS 3.3.1.1, "RPS Instrumentation," a report shall be submitted within 90 days of,nteing CONDITION Fthe following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans to- and schedule for restoring the required instrumentation channels to OPERABLE status."

The new TS section is numbered TS 5.6.8, because on November 21, 2012, NMP2 submitted a License Amendment Request (LAR) to create a new TS section that is numbered TS 5.6.7 for the Reactor Coolant System Pressure and Temperature Limits Report (Reference 5). NMPNS anticipates that LAR will be approved by the NRC and implemented at NMP2 prior to approval of the MELLLA+ LAR. The numbering of TS 5.6.8 is an administrative consideration. The 7 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE MELLLA+ LAR is independent of the LAR submitted on November 21, 2012. NRC approval or rejection of Reference 5 would have no technical impact on the MELLLA+ LAR.

2.3 Modification Summary The MELLLA+ core operating domain expansion does not require major plant hardware modifications. The core operating domain expansion involves changes to the core power/flow map and a small number of setpoints and alarms. Because there are no increases in the operating pressure, power, steam flow rate, and feedwater flow rate, there are no major modifications to other plant equipment.

The stability solution is being changed from Option III to the DSS-CD solution. The DSS-CD solution algorithm, licensing basis, and application procedures are generically described in the DSS-CD LTR (Reference 2), and are applicable to NMP2. The DSS-CD solution uses the same hardware as the current Option III solution. DSS-CD requires a revision to the existing stability solution software.

The boroen 10 enrichmfent in the sodium pentabor-ate solution in the SLS is incr-eased fo

Ž!25 atom pcrccnt to Ž!92 atom perceent. The increase in the boroen 10 enriehment in the scdium pentaborate solution for- the SLS is sufficient to decrease the sodium pentabernte solution velume sterce in tMle SLS storage tank. In a..iti.n, tc I ne SLS pump discharge pressure acceptance criterion is changed to > 1,335 psig. Changes to instrumentati. n setp..ots will be made to accotunt for these changes. The increase in the SLMCPR for two recirculation loops in operation does not require any physical modifications to structures, systems, or components.

3.0 TECHNICAL EVALUATION

3.1 MELLLA+

Attachments 8 and 10 of this Enclosure provide non-proprietary and proprietary versions of the "Safety Analysis Report for Nine Mile Point Unit 2 Maximum Extended Load Line Limit Analysis Plus (MELLLA+ SAR)," NEDO-33576NP and NEDC-33576P, respectively. The MELLLA+ SAR summarizes the results of the significant safety evaluations performed that justify the expansion of the core flow operating domain for NMP2. The changes expand the operating domain in the region of operation with less than rated core flow, but do not increase the licensed power level or the maximum core flow. The expanded operating domain is identified as MELLLA+.

The scope of evaluations required to support the expansion of the core flow operating domain to the MELLLA+ boundary is contained in NEDC-33006P-A, "Maximum Extended Load Line Limit Analysis Plus," referred to as the MELLLA+ LTR (Reference 1). The MELLLA+ SAR provides a systematic disposition of the MELLLA+ LTR subjects applied to NMP2, including performance of plant-specific assessments and confirmation of the applicability of generic assessments to support a MELLLA+ core flow operating domain expansion. The MELLLA+

operating domain expansion is applied as an incremental expansion of the operating boundary without changing the maximum licensed power, maximum core flow, or the current plant vessel dome pressure. The MELLLA+ SAR supports operation of NMP2 at a licensed thermal power of 3,988 MWt with core flow as low as 85% of rated core flow.

8 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE The MELLLA+ core operating domain expansion does not require major plant systems modifications. NMP2 will implement the DSS-CD solution in accordance with the applicable LTRs (References 3 and 4), including the applicable limitations and conditions. Implementation of DSS-CD requires a revision to the existing stability solution software.

The core operating domain expansion involves changes to the operating power/core flow map and changes to a small number of instrument setpoints. Because there are no increases in the operating pressure, power, steam flow rate, and feedwater flow rate, there are no significant effects on the plant systems outside of the Nuclear Steam Supply System (NSSS). There is a potential increase in the steam moisture content at certain times while operating in the MELLLA+ operating domain. The effects of the potential increase in moisture content on plant systems have been evaluated and determined to be acceptable. The MELLLA+ operating domain expansion does not cause additional requirements to be imposed on any of the safety, balance-of-plant, electrical, or auxiliary systems. No changes to the power generation and electrical distribution systems are required as a result of the MELLLA+ operating domain expansion.

This report also addresses applicable limitations and conditions as described in the MELLLA+

LTR SER for the GEH LTR NEDC-33173P, "Applicability of GE Methods to Expanded Operating Domains" (Methods LTR SER) (Reference 4). A complete listing of the applicable LTR SER limitations and conditions and the sections of the MELLLA+ SAR which address them are presented in Appendices A, B, and C of the MELLLA+ SAR.

Only previously NRC-approved or industry-accepted methods were used for the analyses of accidents and transients. Therefore, because the safety analysis methods have been previously addressed, the details of the methods are not presented for review and approval in the MELLLA+

SAR. Also, event and analysis descriptions that are already provided in other licensing reports or the NMP2 Updated Safety Analysis Report (USAR) are not repeated within the MELLLA+ SAR.

Evaluations of the reactor core and fuel performance, reactor coolant and connected systems, engineered safety features, instrumentation and control, electrical power and auxiliary systems, power conversion systems, radwaste systems and radiation sources, reactor safety performance evaluations were performed. The MELLLA+ SAR summarizes the results of the evaluations that justify the MELLLA+ operating domain expansion to a minimum core flow rate of 85% of rated core flow at 100% RTP.

Section 11.3.1 of Attachments 8 and 10 provides a summary of the modifications that will be required to implement the MELLLA+ operating domain, DSS-CD, and the changes to the SLS.

Section 11.3.2 of Attachments 8 and 10 provides a summary of the MELLLA+ issues including a discussion of the MELLLA+ analysis basis, fuel thermal limits, makeup water sources, design basis accidents, challenges to fuel, challenges to the containment, design basis accident radiological consequences, anticipated operational occurrence analyses, combined effects, non-Loss of Coolant Accident (LOCA) radiological release accidents, equipment qualification, balance-of-plant, and environmental consequences.

An assessment of the risk increase, including core damage frequency (CDF) and large early release frequency (LERF) associated with operation in the MELLLA+ operating domain is provided in Attachment 4 of this Enclosure and Section 10.5 of Attachments 8 and 10 of this Enclosure. The estimated risk increase for at-power events due to MELLLA+ is a delta CDF of 1E-8 and delta LERF of 3E-9. This represents a very small risk change in RG 1.174 9 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE (Reference 6). Based on these results, the proposed MELLLA+ operating domain is acceptable on a risk basis.

3.2 DSS-CD The long-term stability solution is being changed from the currently approved Option III solution to DSS-CD. The DSS-CD solution algorithm, licensing basis, and application procedures are generically described in NEDC-33075P-A (Reference 2) and NEDE-33147P-A (Reference 3),

and are applicable to NMP2 including any limitations and conditions associated with their use and approval. Section 2.4 of the MELLLA+ SAR (Attachments 8 and 10) addresses the change to the DSS-CD stability solution. In addition, a complete listing of the required DSS-CD SER and limitations and conditions and the sections of the MELLLA+ SAR which address them is presented in Appendix C of the MELLLA+ SAR, respectively.

DSS-CD is designed to identify the power oscillation upon inception and initiate control rod insertion (scram) to terminate the oscillations prior to any significant amplitude growth exceeding the applicable safety limits. DSS-CD is based on the same hardware design as Option III.

However, it introduces an enhanced detection algorithm that detects the inception of power oscillations and generates an earlier power suppression trip signal exclusively based.n

........ :.... :ri.d eeni*-mati.n r-..gni. . The existing Option III algorithms are retained (with generic setpoints) to provide defense-in-depth protection for unanticipated reactor instability events.

3.3 Standby Liquid Control System Boron 10 En-.hmen-The SLS is described in Section 9.3.5 of the NMP2 USAR. The system provides a backup capability for shutting down the reactor. The SLS is needed only in the event that sufficient control rods cannot be inserted into the reactor core to accomplish shutdown and cooldown in the normal manner. To accomplish this function, the SLS injects a sodium pentaborate solution into the reactor. The SLS consists of a boron solution storage tank, two positive displacement pumps, two explosive valves (provided in parallel for redundancy), and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated water solution is discharged into the RPV through the high pressure core spray sparger.

The specified neutron absorber solution is sodium pentaborate. It is prepared by dissolving granularly-enriched sodium pentaborate in demineralized water (NMP2 USAR 9.3.5.2). The sodium pentaborate solution is discharged radially over the top of the core through the High Pressure Core Spray (HPCS) sparger. The boron absorbs thermal neutrons and thereby terminates the nuclear fission chain reaction in the uranium fuel. The sodium pentaborate also acts as a buffer to maintain the suppression pool pH at or above 7.0 to prevent the re-evolution of iodine, when mixed in the suppression pool following a LOCA accompanied by significant fuel damage (NMP2 USAR Section 9.3.5.1).

3.3.1 Reactor Boron Cold Shutdown Concentration Requirements The reactor boron concentration requirements for achieving cold shutdown (780 parts per million (ppm) natural boron) is not increased for MELLLA+, because there is no change in fuel type and no change to the operating cycle. The total weight of boron-10 required for cold shutdown (including the 25% margin) does change for MELLLA+, because of a conservative increase in the assumed weight of the reactor coolant in the applicable analysis.

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ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE The shutdown margin is calculated for each plant reload and is documented in the Supplemental Reload Licensing Report (SRLR).

Ii ,"b

at'-N1 ** 1* 4 lift IF* Z-- * - - - n- ýL* AI

~u ~nricnmcnr in ~ou:um rcnuwarurc in imronSLflpfl~C 10 CFR 50.62(e)(4) rvauir-es:

"Each boiling water-retor must have a standby liquid contrl system (SCS) withth eapability of injecting into the reacter- prcssurce vessel a bor-ated water- solutiont at such a flo rate, level of bor-on concentfation and borefn 10 isotope enrichmcnt, and aecountfing o r-eactor pressure vessel voluime, that the resulting reaetivity entrl is at least equivalent to that r-esulting from injection of 86 gallons per- minuite of 13 weight perceent sodium pentaborate dceahydrate solutioin at the nAtur-al boroen 10 isotope abundance into a 251 inch inside diameter-r.ea.t. r pressure . ss. l for-a gien

... design..."

The NRC approved licensing topical r-eport NEDE 31096 A (Rcfer-enee 7) provides a method*b which the boront equivalency r-equirement of 10 CFR 50.62(e)(4) can be demonefstrated.

Equation 1 1 of that docuiment was used to demonstfate injectiont capacity equivalency as "elews.

(Q/'86) x (M2511') x (C-'13) x (9/'49.8) 1 1 Where-Q - expeeted SLS flew rate (gpmn)

A425-L mass of water in the rcaetfr ivessel and recircuat~ion system at hot rated conditions (Ibs) for-a 251 inceh diameter- vessel reference plant

-l mass of water in the NMP2 reactor-vessel and r-eeir-eulation system at hot rated eendifiefts Ibs)

- sodium pentaborae solution concentreation (%494)

E boron 10 isotope enrichment (atom percent)

NMP2 is equipped with a 251 inch diameter reactor- vessel (NMP2 USAR Section 15.)

assumptions utilized int the analyses of the changes to the SLS.

Substituting the cuffent values definted in Table I in the above equationt yields:

82.4/86 X 1 X 13.6413 X 25/19.8 - l.27ý--4 Substituting the new values definted in Table 1 into the above equation yields:!

80/86 X 1 X 13.6/13 X 92/19.8 - 4.52> 1 (Galcula*Ien for 13.6 v, 0/_)

This demonstrt*es that the ber*f equivalent control capacity requirement of 10 CFR 50.62(e).4) is met, when the changes to the SLS flow rate and the boront 10 isotope em-iehment are inceluded.

int addition, the control margin ince s. This is due to -nefeasing the boroen 10 enrichment tenn in the equation by a factor- of3.68 (i.e., 92'25 - 3.68).-

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ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE f41re: - T.I-. requre c

.....sm)I I .. - pto nave 1pump

a o A ra..eow at ii ' ... gpml.

Maintaining the TS SR 3. 1.7.7 aeceptance cr-iter-ia for SLC pump flow rate at 41.2 gpmf provides mar~gin with r-espcct to the r~equired flow rate for- AT-WS mitigation. This issue has bee addressed for-euffent eperation via the NMP-NS Correcetive Action System.

As defined in Table 1, the analyzed SLS injection flow Fate is r-educed to 80 gpm flow r-ate o two SLC System pumps. in perati.n to a ount .. for- diluti.n .fe.ts by GEH4 Safety

. identified Communicationt 10 13, Standby Liquid C.ntrol System Dilution Flew, with additional margin.

Table 1 Assumptions rega.rding SLS P.rforman.

Pafametef Units Current Vaulue New-Value Reactor boron

.. n.. ntatin

... for pp 7. 0 cold shutdown (natur-al boront)

Maximuma allowable solutionf V.+% 144. -144 conceentrationt Minimum allowable*% solutio 4-3-A 4-3-6 conccntfatien Solution eoncentration assumed in Y% 4-3 4446 ATWS anal-isi__

Mfinimumn boroen 10 enrichment foi Aom 2- 9-2 ATWS aftaly~fis Designi-SLS pumip flew Faite 813M 4 Minimum SLS pump flow rate as gPm" 44l-2 4442 defined in T-S 3.1.7 Sing pump flow rate (two. pup i pin 92-4 80) 3-.33.3.2 Change in SLS Pump Discharge Pressure Acceptance Criterion TS SR 3.1.7.7 is revised to increase the acceptance criterion for the SLS pump discharge pressure from >_ 1,327 psig to > 1,335 psig. This change is required due to the increase in the peak upper plenum pressure after SLS pump startup to 1,241 pounds per square inch - absolute (psia) as identified in Tables 9-4 and 9-7 of Attachments 8 and 10 of this Enclosure. Currently, the peak upper plenum pressure after SLS pump startup is 1,236 psia. Thus, the ATWS analysis for MELLLA+ establishes a pressure differential of five psi. The SLS pump discharge pressure acceptance criterion in TS SR 3.1.7.7 is increased by eight psig to address the increase in the upper plenum pressure and provide an additional three psi margin.

33.43.3.3 Anticipated Transient without SCRAM Section 9.3.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) provides a summary of the plant-specific analyses of Anticipated Transients Without Scram (ATWS) to demonstrate that the ATWS acceptance criteria are met for operation in the MELLLA+ operating domain. NMP2 meets the ATWS mitigation requirements in 10 CFR 50.62 for an alternate rod insertion (ARI) system, SLS boron injection equivalent to 86 gpm, and automatic RPT logic. The 12 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE plant-specific ATWS analyses take credit for the ATWS-RPT and SLS. However, ARI is not credited.

Section 9.3.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) provides a licensing basis ODYN ATWS analysis that demonstrates that the ATWS acceptance criteria would be met in the event of a NMP2 response to an ATWS event initiated in the MELLLA+

operating domain.

In addition, a plant-specific ATWS analysis was performed at MELLLA+ conditions that assumed operation of a single SLS pump. The analysis and the results are discussed in Section 9.3.1.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure). It concludes: "All ATWS acceptance criteria are met at MELLLA+ conditions with only a single SLS pump operating."

3.3.5 Suppression Pool Buffering In supp.rt of the Alte.ate Source Term (AST-) meth.dology, the SLS also provides suppr.ession.

peel buffer-ing following a LOCA accompanied by significant fuel damage, preventing r-e evelution of iodine from the suppr-ession peol by maintaining the pool pH4 above 7.0. Seetion 9.3.5.1 of the NMP2 USAR requires a sufficient coneentration and quaftity of sodium pentaborate to be available for injcetion into the reaetor vessel to controel pH4 in the suppr-esio pool for 30 days fllwing a DBA LOCAi.

The r1eduetiont int the mi r u :ired solution volume results in a reduction in. the e s solutiont available for- injectionl to maneneppeso ol p14I Ž 7.0 for- 30 days post LOCA.

The minimum" sodium pentabor-ate solution vouei rqie for injection post LOCA for adequate p14 eontrol is 1,065 lons at the limiting t.e., a sodium pentaborate concentration of 13.6%4). The minimum requir-ed tank vollumie at4 al concentration of 13.6 %i reduced frim 4,558.6 gallons to 1,600 gallons. While this does r educe the amount of cxe available solutiont, adequate margin is maintaincd to ensure that the SLS can perform its required AST support funetion.

The proposed bor-on 10 enr-ichment changes do not impaet the capability to achieve and maintain a pH4 above 7.0 in the suppression pool following a LOCA, because the chemnical pr-opeffies n onenetration of the sodium pentaborate solution injected into the suppr-essin pool will remain the same. Given the rduced volume of solution that wil be available, ther will be a two hour r-eduction int the maximum tome available to add boron to the supeso pool to ananp above 7.0 (noeminal time based on low level alann is within 22- housvess ~ttieo h--

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). Are.vie. of the Emergency Operating Procedures confirmed that the sodium pentaborate

U I., ... solutiont "7* ,* t would beI1.-I...--

(.... C' injeeted

within 30 minutesA following

-It% L . .. .+I^ the*. eccunfeftee

l . .of ,i;.LOCA.

The maaximutm 22 hour- timne period proevides a large margin to the minimm requremet for manual operator- action toinec the sodium pentaborate solution of 30 minuatesq. In addition, the Suppression pool pisntepected to drop bielw:7 foseelda.

Section 9.3.5.3 of the NMP2 USAR delineates that ontly one of the two SL.S loops was assume for suppr-ession pool p1H control operation. The proeposed changes to the SLS do not affect th

- -------------------- --- .. ---- - -- --- I--I-- 0~

~~~~0~~ . . ..... ' . ... ... J . . . . . .. . rr I-of3 13 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE

"* "J f J"QI- --

3.3.C n~ige tin 818 Sterage I unit Soluiuon -Volume The pr-opescd bOroni 10 enriehment value allows the minimum solutiont Volumfe stor-ed in theSL sterage tank th be dedreased to 1,530 gallons at a sodium..pentabrate e.ne.ntr..in of 14.4% and 1,600 gallons at a sodium pentablrVate eoinentraion of 13.6%. The mark uip of NMP2 TS Figt 3.1.7 1 promvided infAttaclhment 1 this mfEntlesur delineates thcepreposed hange in the net SLS storage fankl solutioin volume.

The required minimum lumes for-the 13.6 and 14.4 mar solutin vplumres were der p%49 t by deteamininag the m salution veluifne and then increasing the allumne to acotun frei einimum

1) the dead vsluine not pumped in the reator0 that remains in the SLS and 14PCS piping;a

2) instrument accuracy.

The minimum net solution volume for- injection meets all consider-ations for AT-WS boron injection raes, AST suppression pool pH- control, and assures tha the r-eactor- core boron concentr-ation will be greater- than :780 ppm natural boron equivalent 3-.343.3.3 SLS Pump Relief Valve Setpoint Margin The SLS pump relief valve setpoint margin is the difference between the relief valve nominal setpoint and the maximum SLS pump discharge pressure. A margin of 78 psi provides sufficient margin against inadvertent relief valve lifting. The 78 psi is based on an allowance for the relief valve setpoint drift (typically 3% (3% of 1,600 psi = 48 psi)) and SLS pump pressure pulsations (30 psi).

For MELLLA+ operation during the limiting ATWS event, the relief valve setpoint margin is 205.7 psi. This margin is based on a SLS pump relief valve setpoint of 1552 psig (1600 psig -

3% tolerance (i.e., 48 psig)) and subtracting a SLS pump discharge pressure of 1346.3 psig (i.e.,

1552 psig - 1346.3 psig = 205.7 psi). The margin reduces to 175.7 psi if 30 psi for SLS pump pressure pulsations is taken into consideration (i.e., 205.7 psi - 30 psi = 175.7 psi).

3.3.8 Net Positive Suction Head Available (NIPSH1) for SbS P-unqwr The propesed changes include a reduction in the minimum volume for-the SLS storage tank. Ti results inl at eduction in the swaic head available to provide Net Positive Suction Read (NPSH4Yfot the SLS pumps. The calculation that determnines the SLS pumnp NPSNft did not take any cedi for-the staic head abo-ve the SLS storage tank zero level. The mninimum tank level cof:Fespondinfg to the minimum net volume pefmitted by the proposed change to Figure 3.1.7 1 is greater- than three feet above tankl zero.

3.4 Safety Limit Minimum Critical Power Ratio Cycle specific transient analyses are performed to determine the required SLMCPR and the change in Critical Power Ratio (CPR) [ACPR] for specific transients. To ensure that adequate margin is maintained, a design requirement based on a statistical analysis was selected, in that moderate frequency transients caused by a single operator error or equipment malfunction shall be limited such that, considering uncertainties in manufacturing and monitoring the core operating state, at least 99.9% of the fuel rods would be expected to avoid boiling transition. The lowest allowable transient MCPR limit which meets the design requirement is termed the fuel cladding integrity SLMCPR.

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ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE NUREG-0800, Standard Review Plan, Section 4.4, "Thermal and Hydraulic Design," Acceptance Criterion No. I.B, states, in part, that the limiting (minimum) value of CPR is to be established such that at least 99.9% of the fuel rods in the core would not be expected to experience departure from nucleate boiling during normal operation or anticipated operational occurrences.

A cycle specific Operating Limit MCPR (OLMCPR) is established to provide adequate assurance that the fuel cladding integrity SLMCPR is not exceeded for any anticipated operational transients. The OLMCPR is obtained by adding the maximum value of ACPR for the most limiting transient postulated to occur at the plant to the fuel cladding integrity SLMCPR.

3.4.1 Analytical Methods, Standards, Data and Results NMPNS proposes to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops in operation from >_ 1.07 to > 1.09. The proposed change to the SLMCPR value for two recirculation loops in operation is based on an analysis performed by GNF for NMP2 during Cycle 15 operations with MELLLA+ conditions. The GNF report, GNF-0000-0156-7490-RO-P, "GNF Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR," dated August 26, 2013, supports changing the two recirculation loops in operation value of SLMCPR from >_ 1.07 to >_ 1.09, and maintaining the single recirculation loop in operation value of SLMCPR at > 1.09. These values are based on NRC approved methods and procedures. Attachments 9 and 11 of this Enclosure provide non-proprietary and proprietary versions of the GNF report, respectively.

GNF performed the SLMCPR calculation in accordance with Revision 19 of NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel," (Reference 8) using the following NRC-approved methodologies and uncertainties:

" NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations," (August 1999) (Reference 9).

NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations" (August 1999) (Reference 10).

NEDC-32505P-A, "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel,"

(Revision 1, July 1999) (Reference 11).

Section 2.9 of Attachments 9 and 11 of this Enclosure require NMPNS to "provide the current and previous cycle power/flow map in a separate attachment." Figure 1-1 of Attachments 8 and 10 of this Enclosure provide the power/flow operating map for MELLLA+. This will be the power/flow map for NMP2 operations in Cycle 15 following NRC approval of this License Amendment Request. Attachment 5 of this Enclosure provides the NMP2 power/flow operating map for the current operating cycle.

3.4.2 Major Contributors to SLMCPR Change In general, the calculated safety limit is dominated by two key parameters: (1) flatness of the core bundle-by-bundle MCPR distribution, and (2) flatness of the bundle pin-by-pin power/R-Factor distribution. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher calculated SLMCPR. The MCPR Importance Parameter (MIP) measures the 15 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE core bundle-by-bundle MCPR distribution and the R-Factor Importance Parameter (RIP) measures the bundle pin-by-pin power/R-Factor distribution. The impact of the fuel loading pattern on the calculated two recirculation loops in operation SLMCPR has been correlated to the parameter MIPRIP, which combines the MIP and RIP values.

Another factor besides core MCPR distribution or bundle R-factor distribution that significantly impacts the SLMCPR is the expansion of the analysis domain that comes with the initial application of MELLLA+. The rated power / minimum core flow point is analyzed at a lower core flow (than without MELLLA+) using increased uncertainties that tend to increase the SLMCPR. Also, a new point at off-rated power / off-rated flow was analyzed using the increased uncertainties.

Table 3 of the GNF analysis (Attachments 9 and 11 of this Enclosure) presents the MIP and RIP parameters for the previous cycle and the current cycle along with the two recirculation loops in operation SLMCPR estimates using MIPRIP correlations. In addition, Table 3 of the GNF analysis provided in Attachments 9 and 11 presents estimated impacts on the two recirculation loops in operation SLMCPR due to methodology deviations, penalties, and/or uncertainty deviations from approved values. Based on the MIPRIP correlation and any impacts due to deviations from approved values, a final estimated two loops in operation SLMCPR is determined. Section 2.2 of the GNF analysis (Attachments 9 and 11 of this Enclosure) provides a detailed discussion of the items in Table 3 of the GNF analysis (Attachments 9 and 11 of this Enclosure) that result in the increase in the estimated SLMCPR.

3.4.3 Considerations Addressed in the GNF Analysis Regarding R-Factor, Core Flow Rate and Random Effective Tip Reading, and Fuel Axial Power Shape Penalty Section 2.2.1 of the GNF analysis provides a discussion that justifies an increase in the R-Factor uncertainty value. GNF states that it generically increased the GEXL R-Factor uncertainty to account for an increase in channel bow due to the emerging unforeseen phenomena called control blade shadow corrosion-induced channel bow, which is not accounted for in the channel bow uncertainty component of the approved R-Factor uncertainty. NMP2 has experienced control blade shadow corrosion-induced channel bow. Accounting for the control blade shadow corrosion-induced channel bow, the NMP2 Cycle 15 analysis shows an expected channel bow uncertainty which is bounded by the increased GEXL R-Factor uncertainty. Thus, the use of the increased GEXL RFactor uncertainty value adequately accounts for the expected control blade shadow corrosion-induced channel bow for NMP2 Cycle 15.

Section 2.2.2 of the GNF analysis provides a discussion that identifies that the uncertainty values for the core flow rate and the random effective tip reading in the two recirculation loops in operation calculation were conservatively adjusted by using the single recirculation loop in operation uncertainty values. The GNF analysis states the treatment of the core flow and random effective TIP reading uncertainties is based on the assumption that the signal to noise ratio deteriorates as core flow is reduced.

Section 2.4 of the GNF analysis provides a discussion regarding higher uncertainties and non-conservative bases in the GEXL correlations for the various types of axial power shapes. GNF determined that no power shape penalties were required to be applied to the calculated NMP2 Cycle 15 SLMCPR values.

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ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE 3.4.4 Conclusion The proposed change to revise SL 2.1.1.2 by increasing the SLMCPR for two recirculation loops in operation from > 1.07 to > 1.09 is acceptable, and continues to maintain the same level of safety as the current licensing basis.

3.5 NMP2 TS Changes Table 2 defines the affected NMP2 TS, describes the change, and defines the supporting Attachment to this Enclosure that supports the TS Change.

Table 2 - Changes to NMP2 Technical Specifications NMP2 TS Description of the Change Supporting Attachment SL 2.1.1.2 Increase the SLMCPR for two recirculation loops in Attachments 9 and 11 operation from > 1.07 to > 1.09 TS 3.1.7 - Increase the SLS pump discharge pressure from Section 6.5.3 of SR 3.1.7.7 > 1,327 psig to >_1,335 psig Attachments 8 and 10 TS-83.b.7- hinr-easing the sodium pentaberate ber-on0 Seetiont 6.5.1 of SR 3.A.7. enr-iehment rzguir-ement from Ž!25 atom pcr-eent toAttachments 8 and 10 and S92 .tem.peree.

T-S Figure Reducig the minimum net vlume to 1,600 gallt n Section 6.5.1 of 3.i.74i and 1,530 gallens atScdium pentabonuats i ngentrmaiens Attachments 8 and 10 of 13.6% Sad 14.4%, ersperticlhy T-S-Figure lneroasing the COdi ;an pcItabniate brcon 1in Seetio 6.5.1

3. i7 I enriehmcn egunt tfrom Ž!25 atom pereent toŽAttachiments 8 and 10 92acteforPer."it TS 3.3.1.1 The Required Actions for Condition F are modified to: Complies with DSS-

1) Initiate Action to implement the Manual BSP CD LTR Regions defined in the COLR; 2) Implement the Section 2.4 of Automated BSP Scram Region using the modified Attachments 8 and 10 APRM Simulated Thermal Power - High scram setpoints defined in the COLR; and 3) Initiate action in accordance with Specification 5.6.8 TS 3.3.1.1 Condition G is modified to no longer apply in the event Complies with DSS-a Required Action and associated Completion Time of CD LTR Condition F is not met. Section 2.4 of Attachments 8 and 10 TS 3.3. 1.1 New Condition J is added to address the action to take Complies with DSS-in the event a Required Action and associated CD LTR Completion Time of Condition F is not met. Section 2.4 of Attachments 8 and 10 17 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE Table 2 - Changes to NMIP2 Technical Specifications NMP2 TS Description of the Change Supporting Attachment TS 3.3.1.1 New Condition K is added to address the action to take Complies with DSS-in the event a Required Action and associated CD LTR Completion Time of Condition J is not met. Section 2.4 of Attachments 8 and 10 TS SR Correct an editorial error in Note 3 (i.e., ORRM is Editorial correction 3.3.1.1.13 changed to OPRM)

TS SR Eliminate TS SR 3.3.1.1.16 and references to it in TS Complies with DSS-3.3.1.1.16 Table 3.3.1.1-1 CD LTR and TS Section 2.4 of Table Attachments 8 and 10 3.3.1.1-1 TS Table Change the AV for APRM - Flow Biased STP - Section 5.3.1 of 3.3.1.1-1, Upscale from ":50.55W+60.5% RTP and < 115.5% Attachments 8 and 10 Function 2.b RTP" to "< 0.61W + 63.4% RTP and < 115.5% RTP" TS Table Add a new note that requires the Flow Biased Complies with DSS-3.3.1.1-1, Simulated Thermal Power - Upscale scram setpoint to CD LTR Function 2.b be reset to the values defined by the COLR to Section 2.4 of implement the Automated BSP Scram Region in Attachments 8 and 10 accordance with Required Action F--.4F.2 of TS 3.3.1.1 TS Table Add a new note for Function 2.e, OPRM - Upscale, to Complies with DSS-3.3.1.1-1, denote that following implementation of DSS-CD, CD LTR Function 2.e DSS-CD is not required to be armed while in the DSS- Section 2.4 of CD Armed Region during the first reactor startup and Attachments 8 and 10 during the first controlled shutdown that passes completely through the DSS-CD Armed Region.

However, DSS-CD is considered operable and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region TS Table Change the mode of applicability for TS Table 3.3.1.1- Complies with DSS-3.3.1.1-1, 1, Function 2.e, OPRM-Upscale from Mode 1 to > 18% CD LTR Function 2.e RTP. Section 2.4 of Attachments 8 and 10 TS Table Change the allowable value from "As specified in the Complies with DSS-3.3.1.1-1, COLR" to "NA" CD LTR Function 2.e Section 2.4 of Attachments 8 and 10 18 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE Table 2 - Changes to NMP2 Technical Specifications NMP2 TS Description of the Change Supporting Attachment TS.LCO Add a new requirement that pr...i..bModify the LCO Complies with DSS-3.4.1 to prohibit operation in the MELLLA domain or CD LTR MELLLA+ expanded operating domain as defined in the COLR when in operation with a single recirculation Sections 1.2.4 and loop 3.6.3 2.4 of Attachments 8 and 10 address that MELLLA+ is not analyzed for single loop operation In addition, NMP2 does not currently permit single loop operation while in the MELLLA domain, because it is not analyzed.

TS 3.4.1, Add Required Action B.2 to identify that intentional Complies with DSS-Condition B operation in the MELLLA domain or MELLLA+ CD LTR domain as defined in the COLR is prohibited when a recirculation loop is declared "not in operation" due to Sections 1.2.4 and a recirculation loop flow mismatch not within limits 3.6.3 2.4 of Attachments 8 and 10 address that MELLLA+ is not analyzed for single loop operation In addition, NMP2 does not currently permit single loop operation while in the MELLLA domain, because it is not analyzed.

19 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE Table 2 - Changes to NMP2 Technical Specifications NMP2 TS Description of the Change Supporting Attachment TS 5.6.5 Replace the reference to "Reactor Protection System Complies with DSS-Instrumentation Setpoint for the OPRM - Upscale CD LTR Function Allowable Value for Specification 3.3.1.1" Section 2.4 of with a reference to "The Manual Backup Stability Attachments 8 and 10 Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Simulated Thermal Power - High setpoints used in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1."

TS 5.6.8 Add a new TS section (i.e., TS 5.6.8) to define the Complies with DSS-contents of the report required by new Required Action CD LTR F-.2-.F.3 of TS 3.3.1.1 Section 2.4 of Attachments 8 and 10 3.6 TSTF-493 There are no effects on the current TS or their licensing bases relative to TSTF-493. Two TS Reactor Protection System (RPS) functions are changing in this amendment: (1) the OPRM -

Upscale function; and (2) the APRM - Flow Biased Simulated Thermal Power (STP) - Upscale function. The OPRM setpoints are unique to a particular core design for a particular fuel cycle.

The OPRM function setpoints do not have specific TS allowable values (AVs). The APRM STP -

High AVs are specified in TS Table 3.3.1.1-1.

MELLLA+ changes the OPRM setpoints in that they are now derived from DSS-CD algorithms versus Option III algorithms; however, their protective function remains the same. The revised Bases for TS 3.3.1.1 provided in Attachment 2 of this Enclosure states: "The OPRM Upscale function settings are not traditional instrumentation setpoints determined under an instrument setpoint methodology. There is no Allowable Value for this Function, and the OPRM Upscale Function is not [Limiting Safety System Setting (LSSS) Safety Limit (SL)]-related and [the DSS-CD Licensing Topical Report, NEDC-33075P-A] confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology."

MELLLA+ also changes the APRM - Flow Biased Simulated Thermal Power - Upscale AV for two loop operations in the MELLLA+ domain and the APRM - Flow Biased Simulated Thermal Power - Upscale function is used for the Automated Backup Stability Protection (ABSP) if the OPRM becomes inoperable. The APRM STP-High AV and setpoint do have setpoint methodology applied as described in TSTF-493. In addition, the TSTF-493 footnotes were previously added to this function in Amendment 140 to the NMP2 Renewed Operating License NPF-69 issued on December 22, 2011 (Reference 12).

20 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE 3.7 Topics Discussed During NRC Pre-Meetings On February 27, 2013, representatives from NMPNS met with the NRC to discuss the MELLLA+ LAR. During this meeting, the NRC sought clarification regarding several topics.

Table 3 summarizes those topics, and provides a cross reference to the location in Attachments 8 and 10 of this Enclosure that addresses the topic. The NRC issued a summary of this meeting on March 13, 2013 (Reference 13).

Table 3 - Topics Discussed During NRC Pre-Meeting on February 27, 2013 Topic as Summarized in NRC Meeting Summary MELLLA+ SAR Issued on March 13, 2013 (Reference 13) Reference (Attachments 8 and 10 of this Enclosure)

Automated Backup Stability Protection The NMPNS submittal is The NMP2 submittal is based on Revision 6 of NEDC-33075P- based on Revision 7 of A. The NMP2 is planning to take exception to Rev 6 relative to NEDC-33075P-A.

the Automatic Backup Stability Protection (ABSP) set points by using a simplified method that is consistent with the ABSP set Since the February 27, 2013 point methodology described in Revision 7 of NEDC-33075P. meeting, the NRC approved Since the NRC staff has not approved Revision 7 of the Licensing Revision 7 of NEDC-Topical Report (LTR) NEDE-33075P, Re: Detect and Suppress 33075P-A Solution-Confirmation Density (DSS-CD) for Automatic Backup Stability Protection (ABSP), the License Amendment Request Justification provided in (LAR) should not refer to revision 7 of NEDE-33075P, but Section 2.4.3 provide the justifications, consistent with revision 7, for any exceptions taken in the LAR.

Emergency Core Cooling System NPSH Information provided in The NMP2 does not take credit for Containment Accident Section 4.2.6 Pressure (CAP) to assure adequate net positive suction head (NPSH). In response to NRC staff, the licensee stated that a re-analysis of CAP is not required as a result of MELLLA+. Based on feedback from the NRC staff, the NMP2 MELLLA+ submittal will reference the NMP2 Extended Power Uprate (EPU) submittal Requests for Additional Information (RAr's) related to CAP and describe that the NPSH margins in the NMP2 EPU responses remain bounding for MELLLA+.

DSS-CD Implementation Information provided in Implementation of DSS-CD Stability Solution in Place of Option Section 2.4.1 III. The NMP2 MELLLA+ submittal will address the implementation strategy for DSS-CD, including the need for monitoring the timing for arming the protection associated with DSS-CD and the Oscillation Power Range Monitor (OPRM) data analysis already completed.

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ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE Table 3 - Topics Discussed During NRC Pre-Meeting on February 27,2013 Topic as Summarized in NRC Meeting Summary MELLLA+ SAR Issued on March 13, 2013 (Reference 13) Reference (Attachments 8 and 10 of this Enclosure)

TRACG ATWS with Core Instability (ATWSI) Information provided in The NMP2 submittal will include anticipated transients without Section 9.3.3 SCRAM with instability (ATWSI) sensitivity analysis results using a modified T-min correlation similar to what General Electric Hitachi Nuclear Energy (GEH) provided in response to another licensee's RAI. Additional information on the model was requested if and when it becomes available. However, GEH noted that there is no additional testing at this time.

Operator Training Information provided in Provide the implementation plan outlining the simulator upgrade Section 10.6 and operator training plan to support implementation of the LAR.

NMPNS has requested that the NRC approve this LAR by October 2014. To support this schedule, NMPNS plans to upgrade the simulator by the second quarter of 2014 to support operator training in the second and third quarters of 2014.

Reference Core versus Actual Cycle Specific Core See Notes I through 3 Cycle Specific Core Design and Associated Safety Analyses, and Reload Analysis using PRIME Code. The NMP2 submittal will Information provided in describe the potential differences in the analytical inputs and Sections 2.1, 2.2, and 2.6.3 results between the reference core and the actual reload analysis and Footnote 4 of Appendix that will be submitted as a supplement to the MELLLA+ A submittal.

GESTR-M versus PRIME Following the NRC Subsequent to the meeting the NRC staff noted that the licensee's discussions, the MELLLA+

presentation stated that the licensee's LAR submission is going to SAR was revised to utilize include the analyses based in GESTR-M Code and it is planning PRIME Thermal-to supplement its LAR with the Analyses based on PRIME Code, Mechanical (T-M)

The LAR submission based on GESTR-M Code would not be methodology. In addition, "acceptable," This staff concern has been communicated to the PRIME fuel parameters licensee on March 12, 2013. have been used in the analyses requiring fuel In an email dated March 12, 2013, the NRC staff noted that a performance parameters.

LAR submission based on GESTR-M Code would not be "acceptable". A follow-up meeting with the NRC was conducted Information provided in on March 29, 2013. Table 1-1, Sections 2.6.3 and 4.3 22 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE Table 3 - Topics Discussed During NRC Pre-Meeting on February 27, 2013 Topic as Summarized in NRC Meeting Summary MELLLA+ SAR Issued on March 13, 2013 (Reference 13) Reference (Attachments 8 and 10 of this Enclosure)

Notes:

1. The fuel and cycle-dependent analyses, including the plant-specific thermal limits assessment, will be submitted for NRC staff confirmation by supplementing the initial MELLLA+ Safety Analysis Report (SAR) in accordance with Limitation and Condition 12.4 of the MELLLA+ Licensing Topical Report (LTR) Safety Evaluation Report (SER).

Specifically, CENG will provide the cycle specific Supplemental Reload Licensing Report (SRLR) and Fuel Bundle Information Report (FBIR), which includes the supplemental information to satisfy MELLLA+ LTR SER Limitation and Condition 12.4. CENG will submit this information by February 28, 2014.

2. Nine Mile Point Nuclear Station, LLC (NMPNS) will provide a cycle-specific core design loading map along with a summary of differences between the reference design described in the M+SAR and the cycle-specific core design. This summary will include differences in the energy requirements, average enrichment, and analytical inputs, a cycle-specific thermal limits assessment, and the actual reload analysis results. Additionally, the Supplemental Reload Licensing Report, which includes the cycle specific core map, will be provided.

Submittal of the cycle-specific design will satisfy the NRC request made at the MELLLA+

LAR pre-meeting on March 13, 2013.

3. The NMP2 Cycle 15 specific reload analysis will utilize TRACG rather than ODYN for AOO. Section 9.1.1 of the MELLLA+ SAR (Attachments 8 and 10 of this Enclosure) states:

"In the event that the cycle-specific reload analysis is based on TRACG rather ODYN for AOO, no 0.01 added to the OLMCPR is required."

4.0 REGULATORY EVALUATION

4.1 Evaluation of NMIP2 License Amendment Requests to Establish that They Are Not Linked 4.1.1 Guidance from NRR Office Instruction LIC-109, "Acceptance Review Procedures" Revision 1 of Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-109, "Acceptance Review Procedures," (Reference 14) provides the NRR staff (and other NRC staff supporting NRR licensing activities) a basic framework for performing an acceptance review upon receipt of a Requested Licensing Action (RLA) from a licensee. It defines that the NRC should not accept for NRC review and approval an RLA that is linked to another RLA.

Section 1.3.2 of LIC-109 states linked RLAs "are RLAs, where approval of one RLA is contingent upon the approval of (an) other RLA(s) currently under review. This definition evaluates the independence of an RLA with respect to all other RLAs currently under review."

Section 3.1.1 of LIC-109 states: "Linked RLAs: Determine whether the approval of the RLA is contingent upon the approval of other RLAs currently under review. It is important to note that multiple RLAs can affect the same systems or Technical Specifications (TSs) without being 23 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE linked. As such, it may be possible to issue them in any order and without regard to the results of the review of the others. An RLA should not be accepted for NRC review and approval until all prerequisite RLAs have been reviewed and approved by the NRC."

In addition, Example 3 of LIC-109 provides the following example of linked RLAs.

"While the NRC staff is reviewing a licensee's request to change the accident analyses for a loss-of-coolant accident (LOCA), the licensee submits an application for an extended power uprate (EPU). The analysis and supporting justification for the EPU are based, in part, on the proposed LOCA analysis currently under review."

LIC-109 states that this example is not acceptable, "because the EPU should not begin until all prerequisite reviews have been completed (Linked RLAs). Additionally, the regulatory basis cited in the EPU application (i.e., the currently unapproved LOCA analysis) is not the current licensing basis for the plant (Regulatory Basis)." It stated that it may be acceptable for review if "review and approval of the EPU was not contingent upon the outcome of the NRC Staff's review of the LOCA analysis."

4.1.2 Evaluation of NMP2 RCS Pressure - Temperature and MELLLA+ License Amendment Requests On November 21, 2012, NMP2 submitted a License Amendment Request (LAR) to create a new TS section that is numbered TS 5.6.7 for the Reactor Coolant System Pressure and Temperature Limits Report (Reference 5). In the MELLLA+ LAR, a new TS is added that is number TS 5.6.8.

The numbering of TS 5.6.8 is an administrative consideration. The MELLLA+ LAR is independent of the LAR submitted on November 21, 2012. NRC approval or rejection of Reference 5 would have no technical impact on the MELLLA+ LAR.

li *

1"1

i. #.a 11vaiuauon Of rtiir aLO one IVILLLA=Jk- Ueense~ Amendment Kquests On july 5, 2013, NMPNS submitted a request to amend the NMP2 Renewed Operating License (OL) NPF 69 to incr-ease the isotopie cnrichmcnt of boron 10 in the sedium pentaborae selutl used to p..p..e the n.ut..n absorbcr solutioen in the SLS (R*c.rene. 15). This request in:ludes the supporting ehanges to the NMP2 Teehnical Speeification (TS) 3.1.7., "Standby Liquid Control (SLC) System," to increase the boron 10 isotopic enriehmifent in the sodium pentaborate solution utilized in the SLC System and to deer-ease the SLG System tank Yolume.

The SLS LAR and the MELLLA+ LAR both affeet NMP2 TS 3.1.7, including the same changes to SR 3.1.7.10 and Figure 3.1.7 1 to inerease the isotopic enrichment of boroen 10 int the sodium As statedi in Seeiont 1.3. of 11 LI llllnked7V~li~iI 9, RLA "af Rb]s where aproa *i~il o11f* R Is pcntaberate solution and the associated change in the SLS Tank Minimum volume. Section 3.1.1 of LIC 109 establishes that multiple RLAs can affect the same systems or. TSs without bin cntingent upon the approval of (an) other-RLA(s) curffenly under-review.'

of boron 10 in the sodium pentaboa.e solution and reduce the SLS Tank Minimum Yolume requirements, so that they eould be implemented during the spring r-efifeling outage in 2014 for NMP2. These changes will be justified utilizing the current licensing basis, and ae no 24 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE dependent efn the analysis that will be submited in the MELLLA-- LAR. The NRC an review and approve the SLS LAR witheut referenec te the MELLLA+ LR The MELLLA+- LAR propeses changes to the NMP2 T-8, inceluding ehanges to NMP2 T-S31.

that would inrease the SLS pump discharge pressure aceptance . .- iterc-n, incrIease the isotpic enrichment of boroen 10 in the sodium pentaborate solution, and r-educe the SL's Tank Minimumif volume requirements. These changes will be justified utilizing analyses that are m e"d in the MELL.LA+- LAR, inceluding a MELLLA+i specifie boron equivalency analysis and AT-W8 antalysis. Given that the justification for-thesc changes will be proevided int the MELLLA+ LAR, the NRCG can review and approvye the MELLLA+ LAR without refer-ence to the SLS LAR.

if the SLS LAR is approved by the NRC and implemented prior to the NRC approyal of the MELLLA+ LAR, the onily impact to the hMELLLA+ LAR would be to r-emove the proeposed changes to SR 3.1.7.10 and F-igure 3.1.7 1. The antalyses proevided int the MELLLA+ LAR justivf' that those values arc approeprafte for- operatfion in the MELLL6A+ domfain. Thus, those analyses remain valid, and NRC r-evie ireued to justif' operation int the MELLLA+- dom-alin with41 those SLS porametefs.

if the SLS LAR is not approeved by the NRC, this action would have no impact on the NRC-r-eview and approval of the MELLLA+- LAR.

if the MELLLA+ L6AR is approved by the NRC and implemented pr~ior to the NRC approvl o the SLS LAR, then the SLS LAR would be retracted by NMPNS because the SLS [AR does not address oper-ation in the MELLLA+- oper-ating domain and the applicable changes wouldb if the N4BLLLA+/-- LAR is not approved by the NRC, this action would have no impact en the Klt I lllU I+! lill alit a 1 ,1 .,' .. , tin u L Q1t Cs ADv~

N -t&J

-FF 4..44.1.3 NMP2 Conclusion Given the above, the RCS Pressure - Temperature LAR (Reference 5), the SLS LAR (Refer-ence 4--5) and the MELLLA+ LARs are separate and independent licensing actions that the NRC can review and approve independently. Thus, they are not linked RLAs as defined in LIC- 109.

4.2 Applicable Regulatory Requirements/Criteria 4.2.1 MELLLA+ and DSS-CD 10 CFR 50.46 10 CFR 50.46(a)(1)(i) states: "Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section..."

25 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE The acceptance criteria of 10 CFR 50.46(b) are:

"(1) Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 22000 F.

"(2) Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation...

"(3) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

"(4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

"(5) Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core."

Section 4.3 of the MELLLA+ SAR demonstrates that the requirements established in 10 CFR 50.46(a)((1)(i) and the acceptance criteria of 10 CFR 50.46(b)(1) through (5) will be met during operation of NMP2 in the MELLLA+ operating domain.

Appendix A to 10 CFR 50, General Design Criteria 10 CFR 50.36(c)(2)(ii), Criterion 2 requires that TS limiting conditions for operation include process variables, design features, and operating restrictions that are initial conditions of design basis accident analysis. Compliance with the TS ensures that the NMP2 system performance parameters are maintained within the values assumed in the safety analyses. The TS changes are supported by the safety analyses and continue to provide a level of protection comparable to the current TS. Applicable regulatory requirements and significant safety evaluations performed in support of the proposed changes are described in Attachments 8 and 10 of this Enclosure.

Information Notices 2009-23 and 2011-21 NRC Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," notified licensees that analyses performed using pre-1999 methods may be less conservative than previously understood (References 16 and 17). In addition, NRC Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation," notifies addresses that the impact of irradiation on fuel thermal conductivity has the potential to cause errors in ECCS evaluation models, specifically a higher peak cladding temperature (Reference 18).

This issue does not apply to this submittal, because the MELLLA+ SAR utilized to justify operation in the MELLLA+ operating domain includes PRIME T-M methodology as discussed in Section 2.6.3 of Attachments 8 and 10 of this Enclosure.

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ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE 4.2.2 Standby Liquid Control System Isotopic Enrichment of Boron 1 Appendix A to 10 CFR 50, General Design Criteria General Design Criterion (GDC) 26, "Reactivity control system redundancy and capability,"

states:

"Two independent reactivity control systems of different design principles shall be provided.

One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions."

For BWRs, the provisions of 10 CFR 50.62 require that the second reactivity control system be the SLS. Its function is, per the requirements, to inject into the reactor pressure vessel a borated water solution at a prescribed flow rate, concentration and boron-10 isotopic enrichment. The boron in the solution absorbs neutrons, thus providing reactivity control to shut down the reactor in the event the control rods fail to insert into the core.

GDC 27, "Combined reactivity control systems capability," states:

"The reactivity control system shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

The SLS is the poison addition system described in GDC 27.

10 CFR 50.62, "Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants" 10 CFR 50.62 (c)(4) states:

"Each boiling water reactor must have a standby liquid control system (SLCS) with the capability of injecting, into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration, and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the resulting reactivity control is at least equivalent to that resulting from injection of 86 gallons per minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope abundance into a 251-inch inside diameter reactor pressure vessel for a given core design. The SLCS and its injection location must be designed to perform its function in a reliable manner..."

In the NRC-approved licensing topical report, NEDE-31096P-A, "Anticipated Transients Without Scram: Response to NRC ATWS Rule, 10 CFR 50.62," General Electric provides guidance on modifications to the SLC system to ensure licensee compliance with the ATWS rule. The NRC approved the methods presented in NEDE-31096P-A for use by Boiling Water Reactor licensees 27 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE to demonstrate compliance with the ATWS Rule. The application of this guidance demonstrates that the equivalency requirement of 10 CFR 50.62 is met.

10 CFR 50.67, "Accident source term" 10 CFR 50.67.b(1) provides guidance to licensees with respect to revision of the licensee's current accident source term in design basis radiological consequence analyses. Specifically, the regulation states that in order to revise the accident source term, a licensee shall apply for a license amendment under 10 CFR 50.90 and that the application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report.

The radiological consequences of certain design basis accidents (DBA) have been reevaluated using a full implementation of an Alternate Source Term as described in Regulatory Guide (RG) 1.183 (Reference 19) and NRC Standard Review Plan (SRP) 15.0.1 (Reference 20). The evaluation was performed at 120 percent of the original licensed power to bound the effects of future power uprates. The evaluation demonstrates that the calculated offsite exposures and control room doses meet the criteria of 10 CFR 50.67.

The supporting analyses for Alternate Source Tenn assume the pH of the suppression pool is controlled to prevent the re-evolution of iodine following a DBA LOCA. This is accomplished by injecting the SLS solution (i.e., boron solution) following a DBA LOCA to ensure pH is controlled to a value greater than 7.0. Analysis has confirmed that the SLS will continue to maintain suppression pool pH level above 7.0 following a LOCA which involves significant fission product releases.

Information Notice 2001-13 In response to potential non-conservatisms in pressure calculations related to SLS discharge pressure during ATWS scenarios, the NRC issued Information Notice (IN) 2001-13 (Reference 21). IN 2001-13 requested licensees to evaluate relief valve pressure margins on the SLS and confirm to the NRC that the systems remained in compliance with NRC regulations.

NMPNS determined that the concerns identified in IN 2001-13 were applicable to NMP2.

This LAR is proposing to revise NMP2 TS SR 3.1.7.7 by increasing the minimum required NMP2 SLS pump test discharge pressure from 1,327 psig to 1,335 psig, while maintaining adequate margin for relief valve lift.

4.2.3 Safety Limit Minimum Critical Power Ratio 10 CFR 50.36 10 CFR 50.36(c)(1), requires that power reactor facility TS include safety limits for process variables that protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. The fuel cladding is one of the physical barriers that separate the radioactive materials from the environment. The purpose of the SLMCPR is to ensure that specified acceptable fuel design limits (SAFDLs) are not exceeded during steady state operation and analyzed transients. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Fuel cladding perforations can result from thermal stresses, which can occur from reactor operation significantly above design conditions. Since the parameters that 28 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel cladding damage could occur.

The GNF analysis presented in Attachments 9 and 11 of this Enclosure established that NMP2 continues to meet the requirement of 10 CFR 50.36(c)(1) with the increased acceptance criteria for the SLMCPR for two recirculation loops in operation.

Appendix A to 10 CFR 50, General Design Criteria 10 CFR 50.36(c)(2)(ii), Criterion 10 requires:

"The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences."

The fuel cladding must not sustain damage as a result of normal operation and abnormal operational transients. The reactor core safety limits are established to preclude violation of the fuel design criterion such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.

The GNF analysis presented in Attachments 9 and 11 of this Enclosure established that NMP2 continues to meet 10 CFR 50, Appendix A, Criterion 10 with the increased acceptance criteria for the SLMCPR for two recirculation loops in operation.

4.3 Precedent 4.3.1 MELLLA+ and DSS-CD This license amendment request is based on approved GEH license topical reports (References 1 through 4) and their associated Safety Evaluation Reports. The NMP2 application follows the methodologies and limitations of those LTRs and their respective SERs. On January 21, 2010, Monticllo Nuelear Generating Plant submitted a liccmse amendment request to adopt expanded MELLLA+- operating domain; this license amendment request remains under review*b the NRC. (ADAMS I A ession No. ML100280558),

The NRC approved a similar request on March 28, 2014 for the Monticello Nuclear Generator Plant to permit operating in the MELLLA+ operating domain (ADAMS Accession No. ML14035A248 (Reference 22).

4.3.2 Standby Liquid Contro! System 19otepic Enriehment of Boren 10 The NRC has approved a number-of r-equests to inerease the isotopic enrichment of boroen10i the sedium pentaborate utilized to prepar-e the solutiont that is utilized in the SLS. These incelude

Columbia Gcnerating Station issuance of Amendment Re: Increased Boron Concentration in Standby Liquid Control System (TAC No. ME4789), dated May 18, 2011, (ADAM Accession No. MWI 11170370) (Refcr-enc 23) 29 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE This amendment is similar-to the NMP2 proeposed change with r-espeet to the iner-ease in the Isetepie ber-en lenr-ilhment in the s.dium pentabor.e so.ution. utilized in the SLS. For Columbia Generating Station, the bero.n 10 enrichment in.r.ease was from 22 at.m per.ent t 4 pe..eefit,

.;em .

v Qta1lJ~f~l -h - - C+-- JJ Q+tI L1ftt~t I f-+ IUA -JLL S.I 33~lI -V A -I~lA -IZ Dfl .

ClS A Liquid Conitroel System (TAC Nos. MDI 424 and MD)1425), dated Februar-Y 28, 2007-,

(ADAMS8 Accession No. ML070390215) (Reference 2-4)

This amendment is Sim.. ilr to the NM2 proepesed ehage with r.espe. t to the in.rease in th,

.... berf -efr-nriehitment in the sediuA p.ntabor-at÷ solution utilized in the SL'S and the SLS volum,e de.rease. The am.endment alsc rIedu.ed the sodium pentab.fate.en.entratin;*-

ho w eve .. . .. r-r S .

is ...

net.. ..... a e to the. ......' .. . . . . ........... .........ifie concentr-ation.

4.3.3 Safety Limit Minimum Critical Power Ratio The NRC has approved a number of requests to increase the SLMCPR for two recirculation loops in operation that utilized GNF analysis to support the change. These include:

" LaSalle County Station, Unit 2 - Issuance of Amendment No. 192 Regarding Technical Specification Change For Safety Limit Minimum Critical Power Ratio (TAC No. ME9769),

dated February 27, 2013 (ADAMS Accession No. ML13050A637) (Reference 25)

Cooper Nuclear Station - Issuance of Amendment Re: Revision of Technical Specifications -

Safety Limit Minimum Critical Power Ratio (TAC No. ME8853), dated November 9, 2012 (ADAMS Accession No. ML12299A092) (Reference 26) 4.4 Significant Hazards Consideration Nine Mile Point Nuclear Station LLC (NMPNS) is requesting an amendment to Renewed Facility Operating License NPF-69 for Nine Mile Point Unit 2 (NMP2). The proposed amendment includes supporting changes to the NMP2 Technical Specifications (TSs) necessary to: 1) implement the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) expanded operating domain; 2) change the stability solution to the Detect and Suppress Solution -

Confirmation Density (DSS-CD); 3) use the TRACG04 analysis code; and 4) ietIease-th is.topic enr-i.hment of boron 10 in the Standby Liquid Contol System (SLS); and 5) increase the Safety Limit Minimum Critical Power Ratio (SLMCPR) for two recirculation loops in operation.

The proposed changes to the NMP2 TSs:

The following is a list of the proposed changes to the NMP2 TSs:

Revise Safety Limit (SL) 2.1.1.2 by increasing the SLMCPR for two recirculation loops in operation from > 1.07 to > 1.09

Revise the acceptance criterion in TS 3.1.7, "Standby Liquid Control (SLC) System,"

Surveillance Requirement (SR) 3.1.7.7 by increasing the discharge pressure from

> 1,327 pounds per square inch gauge (psig) to > 1,335 psig R. evise the aeceptance cr.iterion in TS SR 3.1.7.10 by increasing the sodium, pentab.- ate boron 10 enrichment r-equir-ement from Ž!25 atom perceent to ý!92 atem perceent, and make-a 30 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE eCorsponding ehange in TS Figure 3.1.7 1, "Sodium PFR cntorPtet Syltien Volum(/P)nsentation R"quircmcnts"

Revise TS Figure 3.1.7 1 to accouint for-the deer-ease in the minimuem volumoe of the SLS tan~k frhm 1,558.6 gallons and 1,288 gallons at sodium pentaborat concentrations of 130/6 and 14.4%a, ospedtively, to 1,600 gallons and 1,530 galTSns at sodium p3ntaborate con1entrations of 13.6%r and 14.4ti, rrespeRtivel

Change the Required Actions for Condition F of TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"

Change Condition G of TS 3.3. 1.1 Add new Conditions lRand K to TS 3.3. 1.1

" Correct an editorial error in Note 3 to TS SR 3.3.1.1.13 (i.e., "ORRM" is changed to "6OPRM"5)

Eliminate TS SR 3.3.1.1.16 and references to it in TS Table 3.3.1.1-1, "Reactor Protection System Instrumentation"

" Change the allowable value (AV) for TS Table 3.3.1.1-1, Function 2.b, Average Power Range Monitor (APRM) - Flow Biased Simulated Thermal Power (STP) - Upscale from

"<5 0.55W +60.5% [Rated Thermal Power] RTP and < 115.5% RTP" to

"< 0.61W + 63.4% RTP and < 115.5% RTP"

" Add a new note to TS Table 3.3.1.1-1, Function 2.b that requires the Flow Biased Simulated Thermal Power - Upscale scram setpoint to be reset to the values defined by the Core Operating Limits Report (COLR) to implement the Automated Backup Stability Protection (BSP) Scram Region in accordance with Required Action F--24F.2 of TS 3.3.1.1

Add a new note to TS Table 3.3.1.1-1, Function 2.e, Oscillation Power Range Monitor (OPRM) - Upscale to denote that following implementation of DSS-CD, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered operable and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region

Change the mode of applicability for TS Table 3.3.1. 1-1, Function 2.e, OPRM-Upscale from Mode 1 to_> 18% RTP

Change the allowable value for TS Table 3.3.1.1-1, Function 2.e from "As specified in the COLR" to "NA"

,Add a prohibition tModify TS Limiting Condition for Operation (LCO) 3.4.1, "Recirculation Loops Operating," the- peibitsto prohibit operation in the Maximum Extended Load Line Limit Analysis (MELLLA) domain or MELLLA+ expanded operating domain as defined in the COLR when in operation with a single recirculation loop

Add Required Action B.2 to TS 3.4.1 to identify that intentional operation in the MELLLA domain or MELLLA+ domain as defined in the COLR is prohibited when a recirculation loop is declared "not in operation" due to a recirculation loop flow mismatch not within limits

" Revise TS 5.6.5.a.4 to replace "Reactor Protection System Instrumentation Setpoint for the OPRM - Upscale Function Allowable Value for Specification 3.3.1.1" with "The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Simulated Thermal Power - High setpoints used in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1"

Add TS 5.6.8, "OPRM Report," to define the contents of the report required by new Required Action R.-F.__3 of TS 3.3.1.1 31 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE NMPNS has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1) Will the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The probability (frequency of occurrence) of Design Basis Accidents occurring is not affected by implementing the MELLLA+ operating domain and DSS-CD stability solution, because NMP2 continues to comply with the regulatory and design basis criteria established for plant equipment.

A SLS failure is not a precursor of any previously evaluated accident in the NMP2 USAR. The increase to the SLMCPR for two recirculation loops in operation does not increase the probability of an evaluated accident. Consequently there is no change in the probability of an accident previously evaluated accident.

The spectrum of postulated transients was investigated and shown to remain within the NRC approved acceptance limits. Fuel integrity is maintained by meeting existing design and regulatory limits. Further, a probabilistic risk assessment demonstrates that the calculated core damage frequency and the large early release frequency do not significantly change due to operation in the MELLLA+ domain.

Challenges to the reactor coolant pressure boundary were evaluated for the MELLLA+ operating domain conditions (pressure, temperature, flow, and radiation) and were found to meet their acceptance criteria for allowable stresses and overpressure margin.

Challenges to the containment were evaluated and the containment and its associated cooling systems continue to meet the current licensing basis. The calculated post LOCA suppression pool temperature remains acceptable.

The SLS is used to mitigate the consequences of an Anticipated Transient Without SCRAM (ATWS) special event and is used to limit the radiological dose during a Loss of Coolant Accident (LOCA). The proposed changes do not affect the capability of the SLS to perform these two functions in accordance with the assumptions of the associated analyses. The ATWS evaluation with the proposed changes incorporated demonstrated that all the ATWS acceptance criteria are met. The ability of the SLS to mitigate radiological dose in the event of a LOCA by maintaining suppression pool pH > 7.0 is not affected by these changes.

This proposed change to the SLMCPR for two recirculation loops in operation does not result in any modification to the design or operation of the systems that are used in mitigation of accidents.

Limits have been established, consistent with NRC approved methods, to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change to the SLMCPR for two recirculation loops in operation continues to conservatively establish this safety limit such that the fuel is protected during normal operation and during any plant transients or anticipated operational occurrences.

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

32 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE

2) Will the change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Equipment that could be affected by implementing the MELLLA+ operating domain and DSS-CD stability solution was evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified. The full spectrum of accident considerations was evaluated and no new or different kind of accident was identified. The MELLLA+ operating domain and DSS-CD stability solution use developed technology and apply it within the capabilities of existing plant safety-related equipment in accordance with the regulatory criteria (including NRC approved codes, standards and methods). No new accident or event precursor was identified.

The long-term stability solution is being changed from the currently approved Option III solution to DSS-CD. DSS-CD is designed to identify the power oscillation upon inception and initiate control rod insertion (scram) to terminate the oscillations prior to any significant amplitude growth exceeding the applicable safety limits. DSS-CD is based on the same hardware design as Option III. However, it introduces an enhanced detection algorithm that detects the inception of power oscillations and generates an earlier power suppression trip signal .. based *n "lusively

.u..e...... p .ei.d .nfirmation rec.gnition. The existing Option III algorithms are retained (with generic setpoints) to provide defense-in-depth protection for unanticipated reactor instability events.

Structures, systems and components (SSCs) previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed changes do not adversely affect safety-related systems or components and do not challenge the performance or integrity of any safety-related system. The physical changes to the SLS a&e-islimited to the incrcase in the boron 10 enrichment of the sodium pentaboratc solution int the SLS8 storage tank, the correspending deerease in the net sodium pentabefmte selution volume roequir-ement in the SLSS storag tank,- the increase in the SLS pump discharge pressure acceptance criterion-aii4,4he ass.ciated inst:um.ntation

.hangcs. The proposed changes do not otherwise affect the design or operation of the SLS.

This proposed change to the SLMCPR for two recirculation loops in operation does not result in any modification to the design or operation of the systems that are used in the mitigation of accidents. The proposed change to the SLMCPR for two recirculation loops in operation assures that safety criteria are maintained.

The proposed changes do not adversely affect any current system interfaces or create any new interfaces that could result in an accident or malfunction of a different kind than was previously evaluated.

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

33 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE

3) Will the change involve a significant reduction in a margin of safety?

Response: No.

The MELLLA+ operating domain affects only design and operational margins. Challenges to the fuel, reactor coolant pressure boundary, and containment were evaluated for the MELLLA+

operating domain conditions. Fuel integrity is maintained by meeting existing design and regulatory limits. The calculated loads on affected SSCs, including the reactor coolant pressure boundary, will remain within their design specifications for design basis event categories. No NRC acceptance criterion is exceeded.

Comprehensive analyses of the proposed changes have concluded that relevant design and safety acceptance criteria will be met without a significant reduction in margins of safety. The analyses have demonstrated that the NMP2 SSCs are capable of safely performing at MELLLA+

conditions. The analyses identified and defined the major input parameters to the Nuclear Steam Supply System (NSSS), analyzed NSSS design transients, and evaluated the capabilities of the NSSS fluid systems, NSSS/Balance of Plant (BOP) interfaces, NSSS control systems, and NSSS and BOP components, as appropriate. Radiological consequences of design basis events remain within regulatory limits and are not increased significantly. The analyses confirmed that NSSS and BOP SSCs are capable of achieving MELLLA+ conditions without significant reduction in margins of safety.

Analyses have shown that the integrity of primary fission product barriers will not be significantly affected as a result of change in the operating domain. Calculated loads on SSCs important to safety have been shown to remain within design allowables with MELLLA+

conditions for all design basis event categories. Plant response to transients and accidents do not result in exceeding acceptance criteria. As appropriate, the evaluations that demonstrate acceptability of MELLLA+ have been performed using methods that have either been reviewed and approved by the NRC staff, or that are in compliance with regulatory review guidance and standards established for maintaining adequate margins of safety. These evaluations demonstrate that there are no significant reductions in the margins of safety.

The SLS is used to mitigate the consequences of an ATWS event and is used to limit the radiological dose during a LOCA. The proposed changes do not affect the capability of the SLS to perform these two functions in accordance with the assumptions of the associated analyses.

The ATWS evaluation with the proposed changes incorporated demonstrated that all the ATWS acceptance criteria are met. The ability of the SLS to mitigate radiological dose in the event of a LOCA by maintaining suppression pool pH > 7.0 is not affected by these changes.

This proposed change to the SLMCPR for two recirculation loops in operation provides a margin of safety by ensuring that no more than 0.1% of fuel rods are expected to be in boiling transition if the MCPR limit is not violated. The proposed change will ensure the appropriate level of fuel protection is maintained. Additionally, operational limits are established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria are met (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences).

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

34 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE 4.5 Conclusions Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Therefore, NMPNS concludes that the proposed amendment presents no significant hazards considerations under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of "no significant hazards consideration" is justified.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does 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 REFERENCES

1. GE Hitachi Nuclear Energy, "General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus Licensing Topical Report," NEDC-33006P-A, Revision 3, June 2009 and NEDO-33006-A, Revision 3, June 2009.

2. GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor, Detect And Suppress Solution - Confirmation Density," NEDC-33075P, Revision 7, June 2011; and Anthony J.

Mendiola (NRC) to Jerald G. Head (GEH), "Revised Draft Safety Evaluation for GE-Hitachi Nuclear Energy Americas, LLC Topical Report NEDC-33075P, Revision 7, 'GE Hitachi Boiling Water Reactor Detect and Suppress Solution - Confirmation Density' (TAC No.

ME6577)," dated August 6, 2013.

3. GE Hitachi Nuclear Energy, "DSS-CD TRACG Application," NEDE-33147P-A, Revision 4, August 2013.

35 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE

4. a. GE Hitachi Nuclear Energy, "Applicability of GE Methods to Expanded Operating Domains," NEDC-33173P-A, Revision 4, November 2012.

b. Letter from R. Kingston (GEH) to NRC, "Clarification of Stability Evaluations - NEDC-33173P," MFN 08-541, June 25, 2008.

c. Letter from J. Harrison (GEH) to NRC, "Implementation of Methods Limitations -

NEDC-33173," MFN 08-693, September 18, 2008.

d. Letter from J. Harrison (GEH) to NRC, "NEDC-33173P - Implementation of Limitation 12," MFN 09-143, February 27, 2009.

e. GE Hitachi Nuclear Energy, "Implementation of PRIME Models and Data in Downstream Methods," NEDO-33173, Supplement 4-A, Revision 1, November 2012.

5. Letter from K. Langdon (NMPNS) to the Document Control Desk (NRC), License Amendment Request Pursuant to 10 CFR 50.90: Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report, dated November 21, 2012 (ADAMS Accession Number ML123380336).

6. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. NRC, Revision 2, May 2011.

7. NEDE 31096 A, "Anticipated Transients Without Senm Respense to NRC ATWS Rl

. .CFR5..62,"Fcbrua, y .. 87..Not used.

8. GE Hitachi Nuclear Energy, "General Electric Standard Application for Reactor Fuel,"

NEDE-2401 1-P-A, and NEDE-2401 1-P-A-US, Revision 19 April 2012.

9 NEDC-32601P-A, "Methodology and Uncertainties for Safety Limit MCPR Evaluations,"

August 1999.

10 NEDC-32694P-A, "Power Distribution Uncertainties for Safety Limit MCPR Evaluations" August 1999.

11 NEDC-32505P-A, "R-Factor Calculation Method for GEl 1, GE12 and GE13 Fuel,"

Revision 1, July 1999.

12. Letter from (NRC) to K. Langdon (NMPNS), Nine Mile Point Nuclear Station, Unit No. 2 -

Issuance of Amendment Re: Extended Power Uprate (TAC No. ME1476), dated December 22, 2011 (ADAMS Accession Number ML1133000041).

13. B. Vaidya (NRC), Summary of February 27,_2013, Meeting with Nine Mile Point Nuclear Station, Unit 2, to discuss Planned Amendment Request on Implementation of MELLLA+

(TAC No. MF0587), dated March 13, 2013 (ADAMS Accession Number ML13059A374)

14. NRR Office Instruction LIC-109, "Acceptance Review Procedures," Revision 1.

36 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE 11; I at..... fxs 12A U..C [Muit

!NWJC tgA D A~n4m~nt I Request Pursuant te 10 CFR 50.90! Standby Liquid Control System Inerease in lsctepie Enrichment of Beroen 10, dated July 1C, 2013 (ADAMS Accessiont Numbef Mb 1 31l9-7A224-l-)Not used.

16. Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," U.S.

Nuclear Regulatory Commission, dated October 8, 2009.

17. Information Notice 2009-23, Supplement 1, "Nuclear Fuel Thermal Conductivity Degradation," U.S. Nuclear Regulatory Commission, dated October 26, 2012.

18. Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation," U.S. Nuclear Regulatory Commission, dated December 13, 2011.

19. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000 (ADAMS Accession No. ML003716792).

20. Standard Review Plan (SRP) 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Revision 0, July 2000 (ADAMS Accession No. ML003721661).

21. Information Notice 2001-13, "Inadequate Standby Liquid Control System Relief Valve Margin," U.S. Nuclear Regulatory Commission, August 10, 2001.

22. Letter from T. Beltz (NRC) to K. Fili (Monticello Nuclear Generating Plant), Moniticello Nuclear Generating Plant - Issuance of Amendment No. 180 to Renewed Facility Operating License repardine Maximum Extended Load Line Limit Anavsis Plus (TAC No. ME3145),

dated March 28, 2014 (ADAMS Accession No. ML14035A248).lcttcr from T. J. O'Connor (Monticcfll Nuclear- Gencrating Plant) to U. S. Nuelear Regulatory Commission, -1iens Amendment Request.: M.. imum Etended Load Line Limit Analysis Pluts," dated Ja"ualry, 21, 2010 (ADAMS Aeecssiefn No. ML100280558).

23. Lcter- from M. C. Thadani (NRC) te M. E. Reddemann (Energy Ncrthwest), "Ccltimbia Gencr-ating Station issuanee of Amendment Re. iner-easd Ber-en ConeentrMien in Standby Liquid Control System (TAC No. ME14789)," dated May 18, 2011 (ADAMS Aeeession No.

Mbin 1 tl-703 70)7Not used.

24. Letter- from R. V. Guizmani (NRC) to 13. T. McKintney (PPL Susquehanna, LLC),

I"Susquehanna A+.. +r.,A A SC Steam

. L T kEkcetric

"-,. uT.. Staion, IV, %units DI.. and DIV "T.' 2kJ Issuance T'-- +. D Amnendment Cf l .+ .. I.- . Standb

-cI '

Liquid Control System (TAG Nos. MD1424 and MD!425)," dated Febrdar-y 28, 200-7 Aeees; Ste" e. r"bt0-t196Y1d:&.t:1"r.1NVt USU,

25. Letter from N. DiFrancesco (NRC) to M. J. Pacillo (Exelon Nuclear), LaSalle County Station, Unit 2 -Issuance of Amendment No. 192 Regarding Technical Specification Change for Safety Limit Minimum Critical Power Ratio (TAC No. ME9769), dated February 27, 2013 (ADAMS Accession No. ML13050A637).

26. Letter from L. E. Wilkins (NRC) to B. J. O'Grady (Nebraska Public Power District), Cooper Nuclear Station -Issuance of Amendment Re: Revision of Technical Specifications -Safety 37 of 38

ENCLOSURE REVISION 1 - EVALUATION OF THE PROPOSED CHANGE Limit Minimum Critical Power Ratio (TAC No. ME8853), dated November 9, 2012 (ADAMS Accession No. ML12299A092).

38 of 38

ATTACHMENT 1 NINE MILE POINT UNIT 2 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS (MARK-UPS)

The current version of the following Technical Specification (TS) page had been marked-up to reflect the proposed changes and replaces the corresponding page previously submitted on November 1, 2013 (ADAMS Accession No. ML13316B107):

3.1.7-3 TS page 3.1.7-4 and associated TS Insert 1 - Figure 3.1.7-1 were implemented with approval of NMP2 Amendment 143, and are to be removed from the original Attachment 1 submitted on November 1, 2013 (ADAMS Accession No. ML13316B107).

The remaining pages in Attachment 1, submitted on November 1, 2013 (ADAMS Accession No. ML13316B107) and modified by the submittal on May 14, 2014 (ADAMS Accession No. ML14139A416), are not changed with this submittal.

Nine Mile Point Nuclear Station, LLC June 13, 2014

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate In accordance

> 41.2 gpm at a discharge pressure with the

> 4-3H psig. Inservice I

Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem 24 months on a from pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.9 Verify all heat traced piping between 24 months storage tank and pump suction valve is unblocked. AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after piping temperature is restored to

> 70°F SR 3.1.7.10 Verify sodium pentaborate enrichment Prior to is > 92 atom percent B-10. addition to SLC tank NMP2 3.1.7-3 Amendment W, 111, 117, 123, 140, 44-37

v t e Letter Sequence Request
TAC:ME3145, Clarify Application of Setpoint Methodology for LSSS Functions (Approved, Closed) TAC:MF2462, Clarify Application of Setpoint Methodology for LSSS Functions (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting, Meeting, Meeting Questions 16 December 2008 11 February 2009 11 March 2009 12 March 2009 18 March 2009 18 March 2009 19 March 2009 20 March 2009 23 March 2009 29 March 2009 28 March 2009 28 March 2009 6 April 2009 6 April 2009 29 April 2009 26 June 2009 2 July 2009 2 July 2009 14 July 2009 14 July 2009 5 November 2009 16 March 2011 12 September 2011 4 October 2012 8 November 2012 28 March 2013 17 May 2013 9 July 2013 26 November 2013 18 December 2008 18 December 2008 19 March 2009 25 October 2009 15 February 2013 12 March 2013 10 May 2013 24 April 2013 Answers 25 January 2010 28 September 2010 13 January 2012 21 December 2012 7 March 2013 18 March 2013 21 March 2013 10 April 2013 18 July 2013 18 July 2013 2 August 2013 18 May 2016 19 July 2012 31 January 2013 29 March 2013 10 April 2013 Results Approval, Approval, Approval, Approval, Approval, Approval Other: L-MT-08-091, Calculation 0801040.301, Steam Dryer Outer Hood Submodel Analysis., L-MT-09-002, Response to NRC Probabilistic Risk Assessment (PRA) Branch Requests for Additional Information (Rais) Dated December 5, 2008, L-MT-09-003, Response to NRC Environmental Branch Requests for Additional Information (Rais) Dated December 18, 2008, L-MT-09-004, Enclosure 1 (Continued) to L-MT-09-004, NSPM Response to Containment & Ventilation Branch RAI Number 2, Copy of MNGP Appr T0406 Data Transmittal Pages 1 of 538 Through Pages 538 of 538, L-MT-09-026, Calculation 0000-0081-6958 MNGP-PRNMS-APRM Calc-2008-NP, Rev. 1, Average Power Range Monitor Selected Prnm Licensing Setpoints - EPU Operation (Numac), L-MT-09-027, Extended Power Uprate: Response to Instrumentation and Controls Branch RAI No. 3 Dated April 6, 2009, L-MT-09-044, Extended Power Uprate: Response to NRC Mechanical and Civil Engineering Review Branch (Emcb) Requests for Additional Information (Rais) Dated March 28, 2009, L-MT-09-049, Drawing C.5-2007, Revision 15, Failure to Scram, L-MT-09-073, Extended Power Uprate: Response to NRC Containment and Ventilation Review Branch (Scvb) Requests for Additional Information (Rais) Dated July 2, 2009 and July 14, 2009, L-MT-09-083, Extended Power Uprate: Limit Curves Requested by the Mechanical and Civil Review Branch (Emcb) Associated with Requests for Additional Information (Rais) Dated March 20, 2009, L-MT-09-088, Extended Power Uprate: Revision to Clarify Text in Enclosures 5 and 7 of L-MT-08-052, L-MT-09-097, Extended Power Uprate: Acknowledgement of NRC Review Delay, L-MT-10-025, Extended Power Uprate (Epu): Response to NRC-NSPM February 25, 2010 Conference Call, L-MT-10-072, Extended Power Uprate: Updates to Docketed Information, L-MT-11-044, Uprate (Epu): Update on EPU Commitments, L-MT-12-056, WCAP-17549-NP, Rev. 0, Monticello Replacement Steam Dryer Structural Evaluation for High-Cycle Acoustic Loads Using ACE, L-MT-12-090, Westinghouse, LTR-A&SA-12-8, Rev. 1, Attachment B, Recommendations for Inspections of the Monticello Replacement Steam Dryer., L-MT-12-114, Drawing C.5-2007, Rev. 17, Failure to Scram, L-MT-13-020, Enclosure 1 - Responses to the Gap Analysis, L-MT-13-029, Enclosure 14 to L-MT-13-029 - WCAP-17716-NP, Revision 0, Benchmarking of the Acoustic Circuit Enhanced Revision 2.0 for the Monticello Steam Dryer Replacement Project., L-MT-13-091, WCAP-17716-NP, Revision 1 - Benchmarking of the Acoustic Circuit Enhanced Revision 2.0 for the Monticello Steam Dryer Replacement Project, Enclosure 14, L-MT-13-092, Extended Power Uprate (Epu): Completion of EPU Commitments, Proposed License Conditions and Revised Power Ascension Test Plan, L-MT-13-096, GE-MNGP-AEP-3304R1, Enclosure 4, NEDC-33435 Corrected Pages, L-MT-13-126, Enclosure 3, GE-MNGP-AEP-3306, Enclosure 2 GEH Response to Mella+ Eicb RAI and Affidavit, L-MT-14-023, Maximum Extended Load Line Limit Analysis Plus: Technical Specification Clarifications, L-MT-15-074, Enclosure 7, WCAP-18604-NP, Revision 0, Monticello EPU Main Steam Line Strain Data Evaluation Report., L-MT-16-017, Revised Commitment to Reconcile Analysis of Bypass Voiding for Transition to Areva Analysis Methodology, L-MT-16-071, Submittal of 2016 Annual Report of Changes in Emergency Core Cooling System Evaluation Models Pursuant to 10 CFR 50.46, ML083230111, ML083400402, ML083500575, ML083570610, ML083590127, ML090710680, ML090710683, ML090710684, ML091120578, ML091140470, ML091410121, ML091410122, ML091410123, ML091410124, ML091760769, ML092290250, ML092810554, ML093160816, ML093220925, ML093220964, ML093620024, ML100280558... further results
MONTHYEAR2009-02-24 [Table View]

ML14169A034

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SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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