ML080840015
| ML080840015 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/26/2008 |
| From: | Wang A B NRC/NRR/ADRO/DORL/LPLIV |
| To: | Entergy Operations |
| Wang, A B, NRR/DORL/LPLIV, 415-1445 | |
| Shared Package | |
| ML080840014 | List: |
| 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
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
- 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
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
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