ML080840015

Issuance of Amendment No. 276, Revise Technical Specification 6.6.5, Core Operating Limits Report, to Add Analytical Methods to Support Implementation of Next Generation Fuel

ML080840015
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/26/2008
From:	Wang A NRC/NRR/ADRO/DORL/LPLIV
To:	Entergy Operations
Wang, A B, NRR/DORL/LPLIV, 415-1445
Shared Package
ML080840014	List: ML080840015 ML080840023
References
TAC MD6220, TAC MD6268
Download: ML080840015 (18)

Text

March 26, 2008

Vice President, Operations

Arkansas Nuclear One

Entergy Operations, Inc.

1448 S.R. 333

Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE:

TECHNICAL SPECIFICATION 6.6.5, "CORE OPERATING LIMITS REPORT (COLR)" (TAC NOS. MD6220 AND MD6268)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 276

to Renewed Facility Operating License No. NPF-6 for Arkansas Nuclear One, Unit No. 2 (ANO-2). The amendment consists of changes to the Technical Specifications (TS) in response

to your application dated July 31, 2007, as supplemented by letters dated July 31, 2007, and March 11, 2008.

The amendment modifies TS 6.6.5, "Core Operat ing Limits Report (COLR)," which would add new analytical methods to support the implementation of Next Generation Fuel (NGF). The

licensee also provided the revised Emergency Core Cooling performance re-analyses in support

of the implementation of Combustion E ngineering (CE) 16x16 NGF as described in WCAP-16500-P-A, "CE 16x16 Next Generation Fuel Core Reference Report." In addition, the

NRC staff approves a one-time application of a 3.5 percent partial credit for NGF thermal margin

gain to be applied to both the Core Operating Limit Supervisory System and Core Protection

Calculator System departure from nucleate boiling calculations (in combination with the

2 percent wide-range penalty) for ANO-2 Cycle 20.

A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be

included in the Commission's next biweekly Federal Register notice. Sincerely,

/RA/

Alan B. Wang, Project Manager

Plant Licensing Branch IV

Division of Operating Reactor Licensing

Office of Nuclear Reactor Regulation

Docket No. 50-368

Enclosures:

1. Amendment No. 276 to NPF-6

2. Safety Evaluation

cc w/encls: See next page 2

Pkg ML080840014, Amdt. ML080840015, License/TS Pgs ML080840023) (*) SE input memo OFFICE NRR/LPL4/PM NRR/LPL4/LA DSS/SNPB/BC DSS/SRXB/BC OGC - NLO w/comments NRR/LPL4/BC NAME AWang (**) JBurkhardt (**) AMendiola (*) GCranston (*) MSmith (**) THiltz DATE 3/25/08 3/25/08 3/6/08 3/26/08 3/26/08 3/26/08 Arkansas Nuclear One (2/25/08)

cc:

Senior Vice President

Entergy Nuclear Operations

P.O. Box 31995

Jackson, MS 39286-1995

Vice President, Oversight

Entergy Nuclear Operations

P.O. Box 31995

Jackson, MS 39286-1995

Senior Manager, Nuclear Safety

& Licensing

Entergy Nuclear Operations

P.O. Box 31995

Jackson, MS 39286-1995

Senior Vice President

& Chief Operating Officer

Entergy Operations, Inc.

P.O. Box 31995

Jackson, MS 39286-1995

Associate General Counsel

Entergy Nuclear Operations

P.O. Box 31995

Jackson, MS 39286-1995

Manager, Licensing

Entergy Operations, Inc.

Arkansas Nuclear One

1448 SR 333

Russellville, AR 72802 Section Chief, Division of Health Radiation Control Section

Arkansas Department of Health and

Human Services

4815 West Markham Street, Slot 30

Little Rock, AR 72205-3867

Section Chief, Division of Health

Emergency Management Section

Arkansas Department of Health and

Human Services

4815 West Markham Street, Slot 30

Little Rock, AR 72205-3867

Pope County Judge

Pope County Courthouse

100 W. Main Street

Russellville, AR 72801

Senior Resident Inspector

U.S. Nuclear Regulatory Commission

P.O. Box 310

London, AR 72847

Regional Administrator, Region IV

U.S. Nuclear Regulatory Commission

611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 276

Renewed License No. NPF-6

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

A. The application for amendment by Entergy Operations, Inc. (the licensee), dated July 31, 2007, as supplemented by letters dated July 31, 2007, and March 11, 2008, complies with the standards and requirements of the Atomic Energy Act of

1954, as amended (the Act), and the Commission's rules and regulations set

forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the

public, and (ii) that such activities will be conducted in compliance with the

Commission's regulations;

D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of

Renewed Facility Operating License No. NPF-6 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through

Amendment No. 276, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical

Specifications

3. The license amendment is effective as of its date of issuance and shall be implemented prior to startup following the spring 2008 refueling outage. Further, Facility Operating

License No. NPF-6 is hereby amended to authorize a change to the Final Safety

Analysis Report (FSAR) to reflect the revised loss-of-coolant accident analyses. The

FSAR changes constitute a change in the analysis of record and will be a baseline for

which future changes will be measured against in accordance with 10 CFR 50.46(a)(3).

This action is required for the implementation of Next Generation Fuel as set forth in the

license amendment application dated July 31, 2007, and evaluated in the safety

evaluation dated March 26, 2008. The licensee shall update the FSAR by adding a

description of this change, as authorized by this amendment, and in accordance with

10 CFR 50.71(e).

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Thomas G. Hiltz, Chief

Plant Licensing Branch IV

Division of Operating Reactor Licensing

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. NPF-6 Technical Specifications

Date of Issuance: March 26, 2008 ATTACHMENT TO LICENSE AMENDMENT NO. 276 RENEWED FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368

Replace the following pages of the Renewed Facility Operating License No. NPF-6 and

Appendix A Technical Specifications with the attached revised pages. The revised pages are

identified by amendment number and contain marginal lines indicating the areas of change.

Operating License REMOVE INSERT

Technical Specifications REMOVE INSERT 6-19 6-19 6-20 6-20 6-21 6-21

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 276 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT NO. 2 DOCKET NO. 50-368

1.0 INTRODUCTION

By application dated July 31, 2007 (Agencywi de Documents Access and Management System (ADAMS) Accession No. ML072200258) (Reference 1), as supplemented by letters dated

July 31, 2007, and March 11, 2008 (ADAMS Accession Nos. ML072200528 and ML080710408, respectively) (References 4 and 3), Entergy Operations, Inc. (Entergy, the licensee), requested

changes to the Technical Specifications (TS) for Arkansas Nuclear One, Unit No. 2 (ANO-2).

The supplemental letters dated July 31, 2007, and March 11, 2008, provided additional

information that clarified the application, did not expand the scope of the application as originally

noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original

proposed no significant hazards consideration determination as published in the Federal Register on August 28, 2007 (72 FR 49576).

The proposed changes would revise TS 6.6.5, "Core Operating Limits Report (COLR)," which

would add new analytical methods to support the implementation of Next Generation Fuel (NGF). The Combustion Engineering (CE) 16x16 NGF design utilizes Optimized ZIRLOŽ, an

advanced cladding alloy. The emergency core c ooling system (ECCS) performance analysis computer codes have been updated to included the Optimized ZIRLOŽ cladding property

changes.

2.0 REGULATORY EVALUATION

The license amendment request involves adding new analytical methods to TS 6.6.5. As

required by paragraph 50.46(a)(1)(i) of Title 10 of the Code of Federal Regulations (10 CFR), the ECCS performance analysis must conform to the ECCS acceptance criteria identified in

10 CFR 50.46(b) which are: Criterion 1, Peak Cladding Temperature - 2200 °F; Criterion 2, Maximum Cladding Oxidation - 0.17 times the total cladding thickness before oxidation; Criterion 3, Maximum Hydrogen Generation - 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; Criterion 4, Coolable Geometry - the

core remains amenable to cooling even calculated changes in core geometry; and Criterion 5, Long-Term Cooling - to maintain an acceptably low calculated core temperature after any

calculated successful initial operation of the ECCS and to remove decay heat for an extended period of time required by the long-lived radioactivity remaining in the core. Additionally, the ECCS performance must be calculated in accordance with an acceptable evaluation model and

must be calculated for a number of postulated loss-of-coolant accidents (LOCAs) of different

sizes, locations, and other properties sufficient to provide assurance that the most severe

postulated LOCAs are calculated. The evaluation model may be either a realistic evaluation

model as described in 10 CFR 50.46(a)(1)(i) or must conform to the required and acceptable

features of Appendix K ECCS Evaluation Models.

Also, regulatory guidance for the review of fuel rod cladding materials and fuel system designs

and adherence to applicable General Design Criteria (GDC) is provided in NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power

Plants," Section 4.2, "Fuel System Design." In accordance with SRP Section 4.2, the objectives

of the fuel system safety review are to provide assurance that:

• The fuel system is not damaged as a result of normal operation and anticipated operational occurrences,

• Fuel system damage is never so severe as to prevent control rod insertion when it is required,

• The number of fuel rod failures is not underestimated for postulated accidents, and

• Coolability is always maintained.

In addition to licensed reload methodologies, an approved mechanical design methodology is

utilized to demonstrate compliance with SRP 4.2 fuel design criteria. The NRC staff has

previously reviewed and approved the CE 16x16 NG F assembly design for application in CE plant designs (Reference 2).

In Section 50.36, "Technical specifications," of Title 10 of the Code of Federal Regulations (10 CFR), the Commission established its regulatory requirements related to the content of TS.

Pursuant to 10 CFR 50.36(d), TS are required to include items in the following five specific

categories related to station operation: (1) sa fety limits, limiting sa fety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements;

(4) design features; and (5) administrative c ontrols. The license amendment request involves adding new analytical methods to TS 6.6.5. The proposed TS changes will be evaluated to

ensure continued compliance with requirements of 10 CFR 50.36(d). Compliance with this

regulation requires a licensee to maintain a list of approved analytical methods (used to

establish potentially cycle-specific core operat ing limits, per NRC Generic Letter (GL) 88-16).

The NRC staff's review will verify that the new analytical methods are applicable to the licensee

and will be used in accordance with established conditions and limitations. This amendment

addresses the requirements for safety limits.

3.0 TECHNICAL EVALUATION

3.1 Proposed Change to Technical Specification 6.6.5

The proposed change to TS 6.6.5, "Core Operating Limits Report (COLR)," involves adding new

analytical methods which will be used to determine the core operating limits related to the

implementation of NGF. The proposed changes are provided in Attachment 2 of Reference 1 with the justification provided in Attachment 1 of Reference 1. The new analytical methods

being added to the COLR, listed below, have been previously reviewed and approved by the NRC:

• WCAP-16500-P-A, CE 16x16 Next Generation Fuel Core Reference Report

• Addendum 1-P-A to WCAP-12610-P-A and CENPD-404-P-A, Optimized ZIRLOŽ

• WCAP-16523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux [CHF] in Rod Bundles with Side-Supported Mixing

Vanes

• CENPD-387-P-A, ABB Critical Heat Flux Correlation for PWR [Pressurized-Water Reactor] Fuel

• Addendum 1-P-A to CENPD-132 Supplement 4-P-A, Calculative Methods for the CE Nuclear Power Large Break LOCA [Loss-of-Coolant Accident] Evaluation

Model - Improvement to 1999 Large Break LOCA EM Steam Cooling Model for Less Than 1 in/sec Core Reflood of Reference 1 identifies and discusses all of the safety evaluation (SE) conditions

and limitations within each of the licensed topical reports (LTRs) being added to COLR. Part of

the disposition includes regulatory commitments stated in Section 4.0 (Attachment 3 of

Reference 1). The NRC staff reviewed the disposition of each SE conditions and limitations and

found, with the exception of WCAP-16500-P-A SE conditions and limitations #5, #6, and #7, that

the licensee adequately addressed each one of them.

3.2 WCAP-16500-P-A SE Conditions and Limitations #5 and #6

WCAP-16500-P-A SE conditions and limitations #5 and #6 deal with the Core Operating Limit

Supervisory System (COLSS) and Core Protec tion Calculator System (CPCS) setpoint methodology and the effects of a mixed core of NGF and non-NGF assemblies. Within

condition and limitation #6, the licensee states that NGF's improved thermal performance offsets any mixed core effects. However, in condition and limitation #5, the licensee states that

a portion of this available thermal margin will be credited by reducing the COLSS and CPCS

addressable constants. Any credit for the NGF assembly thermal characteristics (e.g., mixing

vane CHF correlations) in a mixed core configuration may represent a deviation from approved

methodologies. Sufficient information related to the implementation of this partial credit is not

presented within this amendment request for the staff to determine whether the licensee

complies with approved methods.

In order to resolve these issues, the NRC staff conducted an audit of the Westinghouse

engineering calculations supporting the ANO-2 Cycle 20. This audit was conducted on

February 28, 2008, at the Westinghouse/CE offices in Windsor, Connecticut. During the audit, the NRC staff reviewed the ANO-2 Cycle 20 Core Thermal Hydraulic Reload Analysis (CN-ANO220-010) and the NGF DNB [Departure from Nucleate Boiling] Partial Credit Analysis (CN-TAS-08-6). In addition, the Westinghouse staff presented the methodology used in the

development of a partial thermal margin credit. Based upon the audit, the NRC staff finds that

the partial credit methodology deviates from both the original Modified Statistical Combination of

Uncertainties (MSCU) methodology (CEN-356(V)-P-A) cited in ANO-2 COLR (Reference 3 in TS 6.6.5) and the NGF setpoints methodology (WCAP-16500-P-A) cited in the ANO-2 COLR (Reference 11 in TS 6.6.5).

Recognizing the NGF mixed-core partial credi t as a deviation from previously approved methods, the NRC staff expanded the scope of the audit to review and approve a one-time

methodology deviation for ANO-2 Cycle 20. Us ing the approved TORC model, the partial credit is based on the ratio of iterated heat flux at any given core location over a wide range of

operating conditions.

DNB Adjustment = [ (Minimum Ratio - 1.0 - 0.06)/2 + 1.0 ]

• Minimum Ratio = [ (Heat Flux)Iterated - NGF

/ (Heat Flux)Iterated - CE1

]

• 5 limiting assemblies

• 38 axial shapes

• Rodded and unrodded core operating conditions

• Nominal and off nominal conditions (e.g., Press., Temp., Flow)

• Heat flux iterated to 1.25, 1.0, and 0.8 DNBR [DNB ratio]

The wide range of operating conditions considered in the TORC cases ensures a minimal NGF

thermal margin credit. During the audit, the NRC staff questioned the selection of limiting

assemblies. The five limiting assemblies identified in the Core Thermal-Hydraulics analysis

were all first cycle NGF bundles. The NRC staff had a concern with the application of an NGF

partial credit when the limiting assembly in the core was a non-NGF design. Even though the

current Core Thermal-Hydraulics limiting assembly selection process is not geared toward

minimizing NGF/CE-1 differences, Westinghouse was convinced that the limiting assembly would be an NGF design. Examination of the Cycle 20 loading patterns revealed adjacent first cycle NGF bundles located within low-flow core locations. The higher bundle power, flat radial

power profile, and low-inlet flow combine to yield limiting thermal margin locations. Hence, the

NRC staff concluded it is reasonable to accept Westinghouse's assertion regarding the limiting assemblies.

Furthermore, the calculated minimum ratio using the partial credit methodology was similar in

magnitude to NGF thermal margin gains gat hered from the ANO-2 Cycle 20 CETOP/TORC multipliers (NGF mixed versus CE-1) and in the sample 1/64 hypercube setpoints calculation

within WCAP-16500-P-A (Enclosure 3). This is further proof that the calculated minimum ratio

is reasonable.

The above DNB adjustment equation includes the 6 percent interim margin penalty (WCAP-16500-P-A SE condition #5). Although this margin penalty is not directly applicable to

the partial credit methodology (6 percent based upon the 1/64 hypercube methods), dividing the

net margin gain by a factor of 2.0 yields a conservative partial credit.

For ANO-2 Cycle 20, the minimum heat flux ratios and resulting DNB adjustments are listed

below:

Minimum Ratio (Narrow Range Axial Shape Index (ASI)) = 1.131

Minimum Ratio (Wide Range ASI) = 1.095

DNB Adjustment (Narrow Range ASI) = 1.035

DNB Adjustment (Wide Range ASI) = 1.017

The CPCS algorithms allow for an independent wide-range ASI penalty factor. The current

ANO-2 Reload Data Block constants include a 2 percent wide-range penalty (beyond

+ 0.30 ASI). By crediting the wide-range penalty factors (which are not being credited for any other purpose), a 3.5 percent partial NGF thermal margin gain may be applied to both the

COLSS and CPCS DNB calculations. Although not widely used, credit for the CPCS

wide-range ASI penalty factors is not a methodology change.

In addition to core thermal-hydraulics and COLSS/CPCS setpoints, the NRC staff's audit also

investigated the impact of the mixed core on transient analyses. Westinghouse stated that all

transient thermal margin requirements and accident analyses were performed assuming an

entire core of non-NGF assemblies. CETOP-D and TORC calculations were performed with the

standard assembly dimensions and CE-1 DNB statistics. No credit was taken for the improved

thermal performance of the NGF design.

Ignoring the mixed core may be conservative for many transient analyses (due to ROPM calculations and inherent thermal margin gains with NGF which offset flow starvation).

However, transients dealing with absolute minimum DNBRs (for fuel failure calculations) or fuel

temperature calculations may be adversely affected by specific fuel design characteristics.

During the audit, the NRC staff questioned input and assumptions for the CEA [Control Element

Assembly] Ejection Analysis. It was determined that the analysis correctly included both current

and NGF fuel rod dimensions in the fuel enthalpy calculations. In addition, fuel failure

calculations employed inputs which bounded both current and NGF designs.

COLSS and CPCS addressable constants will be calculated following the approved MSCU

methodology and standard process. CETOP-D calculations were performed with the standard

assembly dimensions and CE-1 DNB statistics. CETOP-to-TORC multipliers were based on a

full core of standard assemblies. Within the statistical analyses, no credit was taken for the

improved thermal performance of the NGF des ign. The partial credit DNB adjustment (discussed above) will be applied directly to the final BERR1 and EPOL2/4 addressable

constants.

Based upon the information reviewed during the audit, the NRC staff finds the one-time

application of the 3.5 percent partial credit (in combination with the 2 percent wide-range

penalty) acceptable for ANO-2 Cycle 20.

3.3 WCAP-16500-P-A SE Condition and Limitation #7

WCAP-16500-P-A SE condition and limitation #7 states that "[i]mplementation of CE 16x16

NGF assemblies necessitate re-analysis of the plant-specific LOCA [Loss of Coolant Accident]

analyses. Licensees are required to submit a license amendment containing the revised LOCA analyses for NRC review. Upon approval, the revised LOCA analyses constitute the analysis of-record and baseline for which future changes will be measured against in

accordance with 10 CFR 50.46(a)(3)." The licensee provided this reanalysis by letter dated

July 31, 2007 (Reference 4). The NRC staff review of this reanalysis is provided below.

3.3.1 Large Break LOCA (LBLOCA)

The Westinghouse ECCS Performance Appendix K Evaluation Model for CE plants is the 1999

Evaluation Model (1999 EM) for LBLOCA.

The 1999 EM for LBLOCA is augmented by CENPD-404-P-A for analysis of ZIRLOŽ cladding and by Addendum 1 to CENPD-404-P-A for

analysis of Optimized ZIRLOŽ cladding. Also, the 1999 EM is supplemented by

WCAP-16072-P-A for implementation of ZrB 2 integral fuel burnable absorber (IFBA) fuel assembly designs.

The 1999EM for LBLOCA includes the following computer codes. The CEFLASH-4A computer

code is used to perform the blowdown hydraulic analysis of the reactor coolant system (RCS) and the COMPERC-II computer code is used to per form the RCS refill/reflood hydraulic analysis and to calculate the containment minimum pressure. It is also used in conjunction with the

methodology described in CENPD-213-P to calculate the FLECHT-based reflood heat transfer

coefficients used in the hot-rod heatup analysis. The HCROSS and PARCH computer codes

are used to calculate steam cooling heat transfer coefficients. The STRIKIN-II computer code is

used for the hot-rod heatup analysis to calculate the peak cladding temperature and maximum

cladding oxidation. Core-wide cladding oxidation is calculated using the COMZIRC computer

code. The initial steady-state fuel rod conditions used in the analysis are determined using the

FATES3B computer code.

The Appendix K steam cooling heat transfer component model for less than 1 inch per second

core reflood in the 1999 EM has been modified to include spacer grid-heat transfer effects. For

ANO-2, the LBLOCA analysis does not credit the use of the modified model including spacer

grid-heat transfer effects.

In performing the LBLOCA calculations, conservative assumptions are made concerning the

availability of safety injection flow. It is assu med that offsite power is lost and all pumps must await diesel startup before they can begin to deliver flow. Also, it is assumed that all safety injection flow delivered to the broken cold leg is lost directly to the containment.

Entergy performed a study to determine the most limiting single failure of ECCS equipment.

The study analyzed no failure, failure of an emergency diesel generator, failure of a

high-pressure safety injection (HPSI) pump, and a failure of a low-pressure safety injection (LPSI) pump consistent with approved topical reports. Maximum safety injection pump flow

rates were used in the no failure case and minimum safety injection pump flow rates were used

in the emergency diesel generator, HPSI ,or LPSI pump failure cases. The pumps were

actuated on a safety injection actuation signal (SIAS) generated by low-pressurizer pressure

with appropriate startup delay. Minimum refueling water storage pool temperature was used in

all four cases as a result of a sensitivity study of the refueling water storage pool water

temperature. The study also investigated the impact of variation in safety injection tank (SIT)

pressure, water temperature and water volume on peak cladding temperature, and peak local

cladding oxidation. A spectrum of guillotine breaks in the reactor coolant pump discharge leg

was analyzed and the results show that the discharge leg is the most limiting break location and

a guillotine break is more limiting than a slot break.

Important core, RCS, ECCS, and containment design data used in the LBLOCA analysis are

listed in Tables 5-1 and 5-2 of Reference 1. The listed fuel rod conditions are for rod average

burnup of the hot rod that produced the highest calculated peak cladding temperature.

Table 5-3 lists the peak cladding temperature and oxidation percentage for the spectrum of

LBLOCAs and times of interest are listed in Table 5-4 of Reference 1. The results of the

full-core implementation of NGF demonstrate conf ormance to the ECCS acceptance criteria as stated in Section 2.0 of this evaluation. These results support a peak linear generation rate of

13.7 kiloWatts per foot.

The NRC staff has reviewed the assumptions, plant design data, and the results of the revised

ECCS performance analysis provided by Entergy and found them acceptable because

conservative assumptions are used and the results (Reference 5) meet the ECCS acceptance

criteria of 10 CFR 50.46(b).

3.3.2 Small Break LOCA (SBLOCA)

The SBLOCA ECCS performance analysis used the Supplement 2 version (referred to as S2M

or Supplement 2 Model of CENPD-137, Supplement 2-P-A, "Calculative Methods for the ABB

CE Small Break LOCA Evaluation Model

"). The S2M for SBLOCA is augmented by CENPD-404-P-A for analysis of ZIRLOŽ cladding, and by Addendum 1 to CENPD-404-P-A for

analysis of Optimized ZIRLOŽ cladding. Also, S2M is supplemented by WCAP-16072-P-A for

implementation of ZrB 2 IFBA fuel assembly designs.

The S2M for SBLOCA uses the following computer codes. The CEFLASH-4AS computer

program is used to perform the hydraulic analysis of the RCS until the time the SITs begin to

inject. After injection from the SITs begins , the COMPERC-II computer program is used to perform the hydraulic analysis. COMPERC-II is only used in the SBLOCA evaluation model for large-break sizes that exhibit prolonged period of SIT flow and significant core voiding. The hot-rod cladding temperature and maximum cladding oxidation are calculated by the STRIKIN-II

computer program during the initial period of forced convection heat transfer and by the PARCH

computer program during the subsequent period of pool boiling heat transfer. Core-wide

cladding oxidation is conservatively represent ed as the rod-average cladding oxidation of the hot rod. The initial steady-state fuel rod conditions used in the analysis are determined using

the FATES3B computer program.

The SBLOCA analysis was performed for the fuel rod conditions that result in the maximum

initial stored energy in the fuel. The calculations included the analysis of both UO 2 and ZrB 2 burnable absorber fuel rods in both the NGF and standard fuel rod designs. For ANO-2, the

analysis was performed using the failure of an emergency diesel generator as the most limiting

single failure of the ECCS. The emergency diesel generator failure causes the loss of a HPSI

pump and LPSI pump, and results in a minimum of safety injection water being available to cool

the core. The LPSI pumps are not explicitly credited in the SBLOCA analysis since the RCS

pressure never decreases below the LPSI pump shutoff head during the portion of the transient

that is analyzed.

A spectrum of three break sizes in the reactor coolant pump discharge (PD) leg was analyzed to

bracket the limiting break size, which for ANO-2 was 0.04 square feet per PD break. The

reactor coolant PD leg is the limiting break location because it maximizes the amount of spillage

from the ECCS. The limiting SBLOCA is the largest small break for which the hot-rod cladding

heatup transient is terminated solely by injection from a HPSI pump.

Important core, RCS, and ECCS design data used in the SBLOCA analysis are listed in

Tables 5-7 and 5-8 of Reference 1. Table 5-9 lists the peak cladding temperature and oxidation

percentages for the spectrum of SBLOCAs and times of interest are listed in Table 5-20 of

Reference 1. The results for the 0.04 square feet per PD break, the limiting SBLOCA, demonstrate conformance to the ECCS acceptance criteria.

The NRC staff has reviewed the assumption, plant design data and the results of the analysis

provided by Entergy and found them acceptable because the assumption is conservative and

the results (Reference 5) meet the ECCS acceptance criteria of 10 CFR 50.46(b). The NRC

staff also agrees with the licensee's conclusion that no SBLOCA mixed-core analysis is

necessary during transition core cycles due to the negligible effect of variation in core hydraulic

losses on SBLOCA analysis results.

3.3.3 Post-LOCA Long-Term Cooling

Entergy stated that the analyses performed with the Westinghouse post-LOCA long-term

cooling evaluation model for CE plants (CENPD-254-P-A) are not sensitive to the fuel assembly changes being introduced for the CE 16x16 NGF design. The NRC staff agrees with the

licensee's conclusion that no plant-specific post-LOCA long-term cooling analyses were

required to support the introduction of the CE 16x16 NGF assembly based on the result of the

analyses performed.

3.3.4 Transition Mixed Core

A transition mixed core assessment was perform ed for NGF in order to address the impact of co-resident hydraulically dissimilar fuel a ssemblies on ECCS performance. The NGF core hydraulic resistance is greater than the standard fuel assembly due to the addition of mixing

grids. Therefore, adjacent NGF and standard assemblies will experience a net redistribution of

flow from the higher resistant NGF assembly to the lower resistant standard assembly.

The flow redistribution in the NGF mixed transition cores produces a slight penalty on the NGF

assembly ECCS performance during the LBLOCA. However, a smaller cross-sectional core

area for coolant flow (relative to a full core of NGF assemblies) is credited in the transition core

assessment to improve the core hydraulics behav ior during the blowdown period. Also, the smaller cross-sectional core area increases the core reflooding rates during the reflood period

relative to the bounding full core of NGF analysis. The net impact on ECCS performance is a

slight reduction in the peak cladding temperature, peak cladding oxidation, and core-wide

cladding oxidation percentage.

For ANO-2, one mixed core configuration was examined to address core loading differences

that are expected in the coming cycles of operation assuming a half core loading pattern for

NGF assemblies. The transition mixed core ECCS performance assessment indicted that the

results were bounded by the results of the full core NGF implementation analysis.

The NRC staff has reviewed Entergy's description for the transition mixed core assessment and

found it acceptable because of the lower impact on ECCS performance.

3.4 WCAP-16500

The NRC staff notes that for any particular cycle-specific core operating limit there are many approved analytical methods which can be used. According to GL 88-16 guidance, TS 6.6.5.b

should list the main approved methods used to support the cycle-specific core operating limit.

Therefore, the NRC staff requested the licensee to identify the main methods to reflect the

GL 88-16 guidance to minimize the number of the approved methods entitled to be listed in

TS 6.6.5.b. In response to the NRC staff request, the licensee committed (Reference 3) to

submit an amendment to minimize the number of references consistent with the guidance

specified in GL 88-16 within 12 months following NRC issuance of the approved amendment for

the current requested changes to TS 6.6.5. This commitment is acceptable because the

TS 6.6.5.b will only list the main methods supporting plant- and cycle-specific operating limits

listed in TS 6.6.5.a. In addition, the NRC staff recommends that the licensee maintain within the

COLR only the current methods being used to determine core operating limits. Superseded

methods should be removed from TS 6.6.5. For example, the COLR lists two NSSS simulation codes, CESEC-III and CENTS. If the CENTS code has replaced CESEC-III for the purpose of

determining core operating limits, then Reference #6 should be deleted.

The NRC staff has reviewed the request by Entergy to approve the revised ECCS analysis to

support the license amendment for the implement ation of CE 16x16 Next Generation Fuel (NGF) described in WCAP-16500 and concludes that the revised ECCS analysis is acceptable

and meets limitation and condition #7 because the analyses used approved methodologies and

results meet the ECCS acceptance criteria of 10 CFR 50.46(b). The NRC staff verified that

each of the LTRs being added to TS 6.6.5 are applicable to the CE-designed ANO-2 reactor

core. Entergy procedure NF-105, "Reload Process," Revision 3 (dated January 22, 2004) provides guidance with regard to development of cycle-specific groundrules. The groundrules document is one of the processes used to assure that LOCA analysis input values bound as-

operated plant values (Reference 6). Based upon compliance with SE limitations and

conditions, the groundrules document, and considering the regulatory commitments identified in

Section 4.0, the NRC staff finds the proposed changes to TS 6.6.5 acceptable. The NRC staff

may audit the licensee in the future to ensure that the licensee's commitments to meet the

requirements specified in the conditions and limitations of each approved methodology is

maintained.

In addition to the changes to TS 6.6.5, the licensee identified administrative changes involving

relocating text within TS pages 6-19 through 6-21. The NRC staff agrees that these changes

are administrative in nature and therefore are acceptable.

4.0 LIST OF REGULATORY COMMITMENTS

The licensee made the following list of regulatory commitments with respect to is its licensing

amendment request. These commitments were identified in Attachment 3 to its application.

1. Additional growth data will be obtained from future LTA [lead test assembly]

exams ahead of the exposure achieved by batch implementation. This data will be provided to the NRC as it becomes available. This is expected sometime

after July 2009 after LTA programs have burnups that bound current ANO-2

burnup limits.

2. If required to maintain acceptable COLSS and CPC DNB operating margin throughout the transition cycle (Cycle 20), a portion of the potential DNB margin

gain may be credited to reduce the DNB uncertainty addressable constants for

COLSS and CPC. In this case even after applying a conservative 6% margin

penalty, no more than one half of the net margin gain, will be credited to reduce

the COLSS and CPC DNB uncertainty addressable constants.

3. The 6% interim margin penalty described in the Limitations and Conditions of the NRC Safety Evaluation of WCAP-16500-P-A will be applied to the resultant

addressable constants until its removal has been approved by the NRC.

4. For the transition cycle, the COLSS on-line monitoring system and the CPC system will continue to utilize the current models and the CE-1 CHF correlation.

5. If the optional steam cooling model described in CENPD-1 32, Supplement 4-P-A, Addendum 1-P and Final Safety Evaluation were to be used for ANO-2

ECCS [Emergency Core Cooling System] Performance Analyses at some time in

the future, then a license amendment request would be submitted including the

analyses and comparison graphical results needed to confirm the acceptability of

the use of the optional steam cooling model.

6. Entergy commits to evaluate other similar plants' TS methodology references that reflect NRC-approved methods used in establishing the COLR parameter

limits. Based on this evaluation, Entergy will propose a change to TS 6.6.5 to minimize the number of references consistent with the guidance provided in GL 88-16, "Removal of Cycle-Specific Parameter Limits from Technical

Specifications." This proposed TS change will be submitted within 12 months

following NRC issuance of the approved amendment for the current requested

changes to TS 6.6.5.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Arkansas State official was notified of the

proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility

component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has

determined that the amendment involves no significant increase in the amounts, and no

significant change in the types, of any effluents that may be released offsite, and that there is no

significant increase in individual or cumulative occupational radiation exposure. The

Commission has previously issued a proposed finding that the amendment involves no

significant hazards consideration, and there has been no public comment on such finding

published in the Federal Register on August 28, 2007 (72 FR 49576). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental im pact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there

is reasonable assurance that the health and safety of the public will not be endangered by

operation in the proposed manner, (2) 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.

8.0 REFERENCES

1. Letter from Entergy to U.S. Nuclear Regulatory Commission, A License Amendment Request to Revise Technical Specification 6.6.5, Core Operating Limits Report," Docket No. 50-368, 2CAN070701, July 31, 2007 (ADAMS Accession No. ML072200258).

2. WCAP-16500-NP-A, "CE 16x16 Next Generation Fuel Core Reference Report" (ADAMS Accession No. ML060670514).

3. Letter from Entergy to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information License Amendment Request to Revise Technical Specification 6.6.5, Core Operating Limits Report," Docket No. 50-368, 2CAN030801, March 11, 2008 (ADAMS Accession No. ML080710408).

4. Letter from Entergy to U.S. Nuclear Regulatory Commission, "Emergency Core Cooling System Performance Analysis," Docket No. 50-368, 2CAN070702, July 31, 2007 (ADAMS Accession No. ML072200528).

5. Letter (2CAN070702) from TGM to USNRC, "Emergency Core Cooling System Performance Analysis," Arkansas Nuclear One, Unit 2, Docket No. 50-368, License No.

NPF-6, July 31, 2007.

5. Letter (2CAN030804) from DEJ to USNRC, 'Response to Request for Additional Information, Emergency Core Cooling System Performance Analysis," Arkansas Nuclear One, Unit 2, Docket No. 50-368, License No. NPF-6, March 20, 2008.

Principal Contributors: P. Clifford T. Huang

Date: March 26, 2008