ML16013A327

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Correction to Amendment No. 205, Maximum Extended Load Line Limit Analysis Plus
ML16013A327
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 02/08/2016
From: Wang A
Plant Licensing Branch IV
To:
Entergy Operations
Wang A
References
CAC MF2798
Download: ML16013A327 (18)
v • d • e


Text

Vice President, Operations Entergy Operations, Inc.

Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 8, 2016

SUBJECT:

GRAND GULF NUCLEAR STATION, UNIT 1 - CORRECTION TO AMENDMENT NO. 205, MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS PLUS (CAC NO. MF2798)

Dear Sir or Madam:

By letter dated August 31, 2015 (Agencywide Documents Access and Management System Accession No. ML15036A258 (not publicly available; proprietary information) and ADAMS Accession No. ML15229A219 (publicly available)), the U.S. Nuclear Regulatory Commission issued Amendment No. 205 to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1 (GGNS). The amendment consisted of changes to the Technical Specifications (TSs) in response to your application dated September 25, 2013, as supplemented by letters dated December 30, 2013; March 10, April 11, July 31, August 14, August 26, September 4, September 10, October 2, October 20, November 20, November 21 (two letters), and December 15, 2014; and January 6, January 20, February 9, February 18, February 19, March 3 and August 13, 2015.

The amendment' revised the GGNS TSs to allow plant operation from the currently licensed Maximum Extended Load Line Limit Analysis (MELLLA) domain to plant operation in the expanded MELLLA Plus (MELLLA+) domain, under the previously approved extended power uprate condition of 4408 megawatts thermal rated core thermal power.

A typographical error was discovered on TS page 3.1-23 of the amendment regarding Surveillance Requirement (SR) 3.1.7.7 of TS 3.1.7, "Standby: Liquid Control (SLC) System Pump Test." The discharge pressure value of 1340 per square inch gauge (psig) was not changed to 1370 psig, as approved in Section 4.2.10.1 of th~ safety evaluation (SE). The NRC also noted that several change bars were inadvertently added on several of the TS pages.

Further, the license pages and TSs "'."ere inadvertently omitted from the proprietary version of the MELLLA+ amendment. As such, the NRC staff is reissuing the three operating license

  • pages and the nine TS pages associated with the MELLLA+ amendment. In addition, Section 10.0 of the SE, page 110, Reference 21, had an incorrect ADAMS accession number.

- 2 ~

A copy of Operating License pages 4, 16f, and 16g; TS pages 3.1-23, 3.3-2a, 3.3-5b, 3.3-6, 3.3-6a, 3.4-1, 5.0-16, 5.0-18, and 5.0-21a; and page 110 of the SE are enclosed. None of these pages contain proprietary information. If you have any questions regarding this matter, please contact me at 301-415-1445.or by e-mail at Alan.Wang@nrc.gov.

Docket No. 50-416

Enclosure:

Sincerely, Alan 8. Wang,, Project Manager Plant Licensing Branch IV-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

1. License and Technical Specifications pages
2. Page 110 of the SE cc w/en~ls: Distribution via Listserv

ENCLOSURE 1 LICENSE AND TECHNICAL SPECIFICATIONS PAGES LICENSE AMENDMENT NO. 205 DATED AUGUST 31, 2015 ENTERGY OPERATIONS, INC.

GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416

(b)

SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above authorized financial transactions, (ii) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS Unit 1 that would materially alter the.

representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment No. 54. In addition, SERI is required to notify the NRC of any action by a lessor or other successor in interest to SE~I that may have an effect on the operation of the facility.

C.

The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in* 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Entergy Operations, Inc. is authorized to operate the facility at reactor core power levels not in excess of 4408 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 205 are hereby incorporated into this license.

Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environm~ntal Protection Plan.

During Cycle 19, GGNS will conduct monitoring of the Oscillation Power Range Monitor (OPRM). During this time, the OPRM Upscale function (Function 2.fof Technical Specification Table 3.3.1.1-1) will be disabled and operated in an "indicate only" mode and technical specification requirements will not apply to this function. During such time, Backup Stability Protection measures will be implemented via GGNS procedures to provide an alternate method to detect and suppress reactor core thermal hydraulic instability oscillations. Once monitoring has been successfully completed, the OPRM Upscale function will be enabled and technical specification requirements will be applied to the function; no further operating with this function in an "indicate only" mode will be conducted.

4 Amendment No. 205

(h)

This license condition shall expire upon satisfaction of the requirements in paragraph (f) provided that a visual inspection of the steam dryer does not reveal any new unacceptable flaw or unacceptable flaw growth that is caused by fatigue.

(47)

Commitments made as required by standard TSTF safety evaluation, as discussed in the notice of availability, will be maintained as described in UFSAR Section 1'6, Technical Specifications. This condition applies to the following TSTFs as approved.

TSTF-423 Changes to the commitments can be made in accordance with 10 CFR 50.59.

(48)

Feedwater Heaters Out-of-Service (FWHOOS)

Operation with FWHOOS in the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) region is prohibited.

(49)

Time Critical Operator Action Commitments made as required fo the MELLLLA+ LAR will be converted to a License Condition as follows:

Prior to Operation in the MELLLA+ Domain, Entergy will:

Train all active operating crews to perform the following three MELLLA+

time-critical operator actions:

1.

Initiate Reactor Water Level Reduction (90 seconds following failure to scram concurrent with no reactor recirculation pumps in service and CTP> 5%).

2.

Initiate Standby Liquid Control Injection (300 seconds if CTP> 5%

or before Suppression Pool Temperature reached 110 degrees 'F).

3.

Initiate Residual Heat Removal Suppression Pool Cooling (660 seconds).

GGNS will validate that all active operating crews have met the time requirements for the three MELLLA+ time-critical operator actions during evaluated scenarios.

GGNS will report any MELLLA+ time-critical actions that are converted to "immediate actions" to the NRC Project Manager.

The Following are one-time actions which expire after the first report:

The results of the three MELLLA+ time-critical operator actions training will be reported to the NRC Project Manager within 60 days of completion of the training.

The reported results will include the full range of response times for each time-critical action and the average times for each crew.

Any MELLLA+ time-critical operator training failures during evaluated scenarios will be reported to the NRC within 60 days of any failures with a plan for resolution.

16f Amendment No. 170, 178, 186, 191, 192,

~.~205

D.

The facility required exemptions from certain requirements of Appendices A and J to 1 O CFR Part 50 and from certain requirements of 1 O CFR Part 100. These include: (a) exemption from General.Design Criterion 17 of Appendix A until startup following the first refueling outage, for ( 1) the emergency override of the test mode for the Division 3 diesel engine, (2) the second level undervoltage protection for the Division 3 diesel engine, and (~)the generator ground over current trip function for the Division 1 and 2 diesel generators (Section 8.3.1 of SSER #7) and (b) exemption from the requirements of Paragraph lll.D.2(b)(ii) of Appendix J for the containment airlock testing following normal door opening when containment integrity is not required (Section 6.2.6 of SSER #7). These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. In addition, by exemption dated December 20, 1986, the Commission exempted licensees from 10 CFR 100.11(a)(1 ), insofar as it incorporates the definition of exclusion area in 10 CFR 100.3(a), until April 30, 1987 regarding demonstration of authority to control all activities within the exclusion area (safety evaluation accompanying Amendment No. 27 to License (NPF-29). This exemption is authorized by law, and will not present an undue risk to the public health and safety, and is consistent with the common defense and security. In addition, special circumstances have been found justifying the exemption. Therefore, these exemptions are hereby granted pursuant to 10 CFR 50. 12. With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act and the rules and regulations of the Commission.

E.

The licensee shall fully implement and maintain in effect all provision of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51FR27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Physical Security, Safeguards Contingency and Training and Qualification Plan," and were submitted to the NRC on May 18, 2006.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensee's CSP vitas approved by License Amendment No. 186 as supplemented by a change approved by License Amendment Nos. 192 and 200.

16g Amendment No. 205

. I

SURVEILLANCE REQUIREMENTS (continued)

SR 3.1.7.6 SR 3.1.7.7 SR 3.1.7.8 SR 3.1.7.9 SR 3.1.7.10 GRAND GULF SURVEILLANCE Verify each SLC subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position, or can be aligned to the correct position.

Verify each pump develops a flow rate

~ 41.2 gpm at a discharge pressure

~ 1370 psig.

Verify flow through one SLC subsystem from pump into reactor pressure vessel.

Determine Boron-10 enrichment in atom percent (E).

Verify piping between the storage tank and the pump suction is not blocked.

3.1-23 Next page is 3.1-26.

SLC System 3.1.7 FREQUENCY 31 days In accordance with the Inservice Testing Program 24 months on a STAGGERED TEST BASIS Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after boron is added to the solution.

Once withJ.n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after solution temperature is restored to

> 45°F Amendment No. -+/--9-fl., -+/--9-+, 205

ACTIONS (continued)

CONDITION I.

As required by Required Action D.l and referenced in Table 3.3.1.1-1.

J.

As required by Required Action D.l and referenced in Table 3.3.1.1-1.

K.

Required Action and associated Completion Time of Condition J not met.

L.

Required Action and associated Completion Time of Condition K not met.

GRAND GULF I. 1 J.1 AND J.2 AND J.3 K. l AND REQUIRED ACTION Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Initiate action to implement the Manual BSP Regions defined in the COLR.

Implement the Automated BSP Scram Region using the modified APRM Flow Biased Simulated Thermal Power -

High trip function setpoints defined in the COLR.

Initiate action to submit an OPRM report in accordance with Specification 5.6.7.

Initiate action to implement the Manual BSP Regions defined in the COLR.

K.2 Reduce operation to below the BSP Boundary defined in the COLR.

AND K.3 L. l

NOTE --------

LCO 3.0.4 is not applicable.

Restore required channels to OPERABLE.

Reduce THERMAL POWER to< 16.8% RTP.

RPS Instrumentation 3.3.1.1 COMPLETION TIME Immeaiate_j_y Immediately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Immediately Immediately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 120 days 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 3.3-2a Amendment No. -+/--&&, -+/--9--+/-,. 205

SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.1.20 SR 3.3.1.1.21 SR 3.3.1.1.22 SURVEILLANCE

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

1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
2. For Functions 2.a, 2.b, and 2.c, the APRM/OPRM channels and the 2-0ut-Of-4 Voter channels are included in the CHANNEL FUNCTIONAL TEST.
3. For Functions 2.d and 2.f, the APRM/OPRM channels and the 2-0ut-Of-4 Voter channels plus the flow input function, excluding the flow transmitters, are included in the CHANNEL FUNCTIONAL TEST.

Perform CHANNEL FUNCTIONAL TEST.

Perform LOGIC SYSTEM FUNCTIONAL.TEST.

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

For Function 2.e, "n" equals 8 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

Testing APRM and OPRM outputs shall alternate.

  • Verify the RPS RESPONSE TIME is within limits.

DELETED RPS Instrumentation 3.3.1.1 FREQUENCY 184 days 24 months 24 months on a STAGGERED TEST BASIS DELETED GRAND GULF 3.3-Sb Amendment No..:t-B-8,- +/--9-1-, 205

FUNCTION

1.

Intermediate Range Monitors

a.

Neutron Flux -

High

b.

Inop RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 4)

Reactor Protection System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS 2

2 5 (a)

REQUIRED CHANNELS PER TRIP SYSTEM 3

3 3

3 CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 H

I H

I SURVEILLANCE REQUIREMENTS SR 3. 3. 1. 1. 1 SR 3. 3. 1. 1. 3 SR 3.3.1.1.12 SR 3.3.1.1.13 SR 3. 3. 1. 1. 1 SR 3. 3. 1. 1. 4 SR 3.3.1.1.12 SR 3. 3. 1. 1. 13 SR 3. 3. 1. 1. 3 SR 3.3.1.1.13 SR 3. 3. 1. 1. 4 SR 3. 3. 1. 1. 13 ALLOWABLE VALUE

$ 122/125 divisions of full. scale

$ 122/125 divisions of full scale NA NA

2.

Average Power Range Monitors

a. Neutron Flux -
High, Set down
b.

Fixed Neutron Flux -

High

c.

Inop

d.

Flow Biased Simulated Thermal Power -

High

e. 2-0ut-Of-4 Voter
f.

OPRM Upscale GRAND GULF 2

1 1,2 1

1,2

~ 16.8%

RTP 3 (c) 3 (c) 3 (c) 3 (c) 2 3 (C) 3.3-6 H

G H

G H

J SR 3.3.1.1.7

$ 20% RTP SR 3.3.1.1.lO(d) (e)

SR 3. 3. 1. 1. 19 SR 3. 3. 1. 1. 2 0 SR 3.3.1.1.2

$ 119.3% RTP SR 3.3.1.1.7 SR 3.3.1.1.lO(d) (e)

SR 3. 3. 1. 1. 19 SR 3. 3. 1. 1. 2 0 SR 3.3.1.1.20 NA SR 3.3.1.1.2 (b) (g)

SR 3.3.1.1.7 SR 3.3.1.1.lO(d) (e)

SR 3. 3. 1. 1. 1 7 SR 3.3.1.1.19 SR 3. 3. 1. 1. 2 0 SR 3. 3. 1. 1. 19 SR 3. 3. 1. 1. 2 0 SR 3.3.1.1.21 NA SR 3.3.1.1.22 SR 3.J.1.1.7 SR 3.3.1.1.lO(d) (e)

SR 3.3.1.1.19 (f)

SR 3. 3. 1. 1. 2 0 (continued)

Amendment No. +/--6-9-, M.&,- +/--9-+/-,* 205

Table 3.3.1.1-1 (page 2 of 4)

Reactor Protection System Instrumentation RPS Instrumentation 3.3.1.1 (a)

With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b)

Two-Loop Operation: 0.64W + 61.8% RTP and~ 113% RTP Single-Loop Operation: 0.58W + 37.4% RTP (c)

Each channel provides inputs to both trip systems.

(d). If the as-found channel setpoint is outside its pre-defined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e)

The instrument channel setpojnt shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the NTSP are acceptable provided the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures to confirm channel performance.

The NTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Technical Requirements Manual.

(f)

The setpoint for the OPRM Upscale Confirmation Density Algorithm (CDA) is specified in the COLR.

(g)

With the OPRM Upscale trip function (Function 2.f) inoperable, reset the APRM Flow Biased Simulated Thermal Power -

High trip function (Function 2.d)setpoints to the values defined\\by the COLR to implement the Automated BSP Scram Region in accordance with Action J of this specification.

GRAND GULF 3.3-6a Amendment No. +/--&-&,- :+/--9-1-, 205

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO

3.4.1 APPLICABILITY

ACTIONS Two recirculation loops with matched flows shall be in operation.

One recirculation loop shall be in operation provided the plant is not operating in the MELLLA+ domain defined in the COLR and provided the required limits are modified for single loop operation as specified in the COLR.

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

Requi red limit modifications for single-recirculation loop operation may be delayed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after transition from two recirculation loop operation to single recirculation loop operation.

MODES 1 and 2.

CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation loop jet A.1 pump flow mismatch not within limits.

GRAND GULF Shutdown one recirculation loop.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (continued)'

3.4-1 Amendment No. -rrB, f-4-+/-, 205

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 5.5.12 GRAND GULF Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1.

A change in the TS incorporated in the license; or

2.

A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes that do not meet the criteria of either Specification 5.5.11.b.1 or Specification 5.5.11.b.2 above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

10 CFR 50, Appendix J, Testing Program This program establishes the leakage rate testing program of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be implemented in accordance with the Safety Evaluation issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087) as modified by the Safety Evaluation issued for Amendment No. 135 to the Operating License, except that the next Type A test performed after the November 24, 1993 Type A test shall be performed no later than November 23, 2008.

Consistent with standard scheduling practices for Technical Specifications required stirveillances, intervals for the recommended surveillance frequency for Type A, B and C testing may be extended by up to 25 percent of the test interval, not to exceed 15 months.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 12.1 psig.

Amendment No. -+/--§.+, ~ 9-+/--, 205

Reporting Requirements 5.6 5.6 *Reporting Requirements 5.6.2 5.6.3

5. 6. 4.

5.6.5 GRAND GULF Annual Radiological Environmental Operating Report (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted in a supplementary report as soon as possible.

Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year.

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be cdnsistent with the objectives outlined in the ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.l.

Deleted Core Operating Limits Report (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1)

LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR) I

2)

LCO 3.2.2, Minimum Critical Power Ratio (MCPR),

3)

LCO 3.2.3, Linear Heat Generation Rate (LHGR),

4)

Deleted

5)

LCO 3.3.1.1, RPS Instrumentation, Table 3.3.1.1-1 APRM Function 2.f

6)

The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region II), the modified APRM Flow Biased Simulated Thermal Power -

High trip function (Function 2.d) setpoints used in the OPRM Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1.

(continued) 5.0-18 Amendment No. H+, +&7, 8&, 205

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 5.6.7 GRAND GOLF Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

a.

RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

i)

Limiting Conditions for Operations Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits" ii)

Surveillance Requirements Section 3.4.11, "RCS Pressure and Temperature (P/T) Limits"

b.

The analytical methods used to determine the RCS pressure a.nd temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

i)

NEDC-33178P-A, "GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Temperature Curves" Revision 1, June 2009

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Oscillation Power Range Monitor (OPRM) Report When an OPRM report is required by CONDITION J*of LCO 3.3.1.1, "RPS Instrumentation," it shall be submitted within the following 90 days.

The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABLE status.

5.0-21a Amendment No. -+/--9-+/-, 205

ENCLOSURE 2 CORRECTED PAGE 110 OF THE SAFETY EVALUATION LICENSE AMENDMENT NO. 205 DATED AUGUST 31, 2015 ENTERGY OPERATIONS, INC.

GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416

OFFICIAL USE ONLY PROPRIETARY INFORMATION

- 110 -

Nuclear Station, Unit 1, Docket No. 50-416, License No. NPF-29," dated February 19, 2015 (ADAMS Accession No. ML15050A077).

21.

Mulligan, K., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Revision to letter GNR0-2015/00013 to Change License Condition Number 48 to 49 and to add License Condition 48 from letter GNR0-2013/00012, Grand Gulf Nuclear Station, Unit 1, Docket No. 50-416, License No. NPF-29," dated March 3, 2015 (ADAMS Accession No. ML15063A075).

22.

Eisenhut, D. G., U.S. Nuclear Regulatory Commission, letter to J.B. Richard, Mississippi Power & Light Company, "Issuance of Facility Operating License NPF 29 - Grand Gulf Nuclear Station, Unit,: November 1, 1984 (ADAMS Accession No. ML021410165).

23 Jaffe, D. H., U.S. Nuclear Regulatory Commission, letter to William A. Eaton, Entergy Operations, Inc., "Grand Gulf Nuclear Station, Issuance of Amendment RE: 1.7%

Increase in Licensed Power Level (TAC No. MB3972)," October 10, 2002 (ADAMS Accession No. ML022630304).

24.

Wang, A. B., U.S. Nuclear Regulatory Commission, letter to Entergy Operations, Inc.,

"Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment RE: Extended Power

25.

De Young, R. C., U.S. Nuclear Regulatory Commission, letter to N. L. Stampley, Mi~sissippi Power & Light Company, Regarding Issuance of Construction Permits.

CPPR-118 and CPPR-119 authorizing construction of the Grand Gulf Nuclear Station, Units 1and2, dated September 4, *1974 (ADAMS Accession No... ML021410145).

26.

U.S. Nuclear Regulatory Commission, "Part 50, Licensing of Production and Utilization Facilities - General Design Criteria for Nuclear Power Plants," Federal Register, Vol. 36, No. 35, February 20, 1971, pp. 3255-3260.

27.

Grand Gulf Nuclear Station, Unit 1, Updated Final Safety Analysis Report," dated September 4, 2014. (Not publicly available).

28.

GE-Hitachi, NEDC-33612P, Revision 0, "Safety Analysis Report for Grand Gulf Nuclear Station Maximum Extended Load Line Limit Analysis Plus," September 2013 (ADAMS Accession No. ML13269A140).

29.

U.S. Nuclear Regulatory Commission, "Appendix K - Safety Evaluation of Supplement 3 to NEDC-33173P, Final Safety Evaluation by the Office of Nuclear Reactor Regulation NEDC-33173P, Supplement 3, Applicability of GE Methods to Expanded Operating Domains - Supplement for GNF2 Fuel," GE-Hitachi Nuclear Energy Americas, LLC Project No. 710," December 28, 2010 (ADAMS Accession No. ML103270383).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

MLML16013A327 Rids Rgn4Mai1Center Resource RidsNrrDraAphb Resource RidsNrrDssStsb Resource OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA N RR/DOR L/LPL4/BC NRR/DORL/LPL4/PM NAME A Wang PBlechman MKhanna AWang DATE 2/5/16 2/4/16 2/5/16 2/8/16

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