ML060620368

Attachment 3 (Technical Specifications) and 4 (Affidavit and Non-Proprietary Version of Attachment 2 to Exelon Letter Dated January 26, 2006)

ML060620368
Person / Time
Site:	Dresden, Quad Cities
Issue date:	01/26/2006
From:	Exelon Generation Co, Exelon Nuclear
To:	Office of Nuclear Reactor Regulation
References
CAW-06-2095, RS-06-009, TAC MC7323, TAC MC7324, TAC MC7325, TAC MC7326 NF-BEX-06-15 NP
Download: ML060620368 (55)

Text

ATTACHMENT 3 Retyped Technical Specifications Page for Proposed Change QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-3Q REVISED TECHNICAL SPECIFICATIONS PAGE 5.6- 4

Reporting Requirements 5 .6 Reporting Requirements

.6 .5 CORE OPERATING LIMITS REPORT (COLR) 10 Fuels Corporation Critical Power for Boiling Water Reactors/Advanced Nuclear ion Critical Power Methodology for Boi I i Reactors : Methodology for Analysis of Assembly el Bowing Effects/NRC Correspondence, ANF-524040 .

11 . COTRANSA 2 : A Computer Program for Boiling Water Reactor Transient Analyses, ANF-913(P)(A) .

12 . Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A) .

13 . Commonwealth Edison Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods ."

14 . ANFB Critical Power Correlation Application for Coresident Fuel, EY-1125(P)(A) .

15 . EMF-85-74(P), RODEX2A(BWR) Fuel Rod Thermal Mechanical Evaluation Model, Supplement 1(P)(A) and Supplement 2 (P)(A), Siemens Power Corporation, February 1998 .

16 . NEDC-32981P, "GEXL96 Correction for ATRIUM 9B Fuel ."

17 . CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel ."

18 . WCAP-16081-P-A, "1000 SVEA Fuel Critical Power Experiments and CPR Correlation : SVEA-96 Optimal ."

19 . WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model : Supplement 2 to Code Description, Qualification and Application ."

20 . WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model : Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optimal Fuel ."

21 . WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling I

Water Reactors - Supplement ."

{continued)

Quad Cities 1 and 2 5 .6-4 Amendment No .

ATTACHMENT 4 Westinghouse Application for Withholding, Affidavit, and Non-Proprietary Version of Attachment 2

Westinghouse Westinghouse Electric Company Nuclear Services P .O . Box 355 Pittsburgh, Pennsylvania 15230-0355 USA latory Commission Direct tel : (412) 374-4419 Document Control Desk Directfax : (412) 374-4011 gtan, DC 20555-0001 e-mail : maurerbfCwestinghouse.com Our ref: CAW-06-2095 January 25, 2006 PLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Westinghouse Input to Dresden Nuclear Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2 - Request for Additional Information Regarding Transition to Westinghouse SVEA-96 Optima2 Fuel (Proprietary/Non-Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-06-2095 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations .

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Nuclear.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-06-2095 and should be addressed to B. F. Maurer, Acting Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O . Box 355, Pittsburgh, Pennsylvania 15230-0355 .

B. F. Maurer, Acting Manager Regulatory Compliance and Plant Licensing Enclosures cc : F. M. Akstulewicz/NRR I. Clifford/NRR M. BanerieefNRR G. S. Shukla/NRR L. M. Feizollahi/NRR (affidavit only)

A BNFL Group company

CAW-06-2095 AFFIDAVIT ONWEALTH OF PENNSYLVANIA :

ss COUNTY OF ALLEGH Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

. Gresham, Manager Regulatory Compliance and Plant Licensing Sworn to and subscribed Notary Public

2 CAW-06-2095 I am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule ma roeeedings, and am authorized to apply for its withholding on behalf of Westin e.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld .

(1) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse .

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public . Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence . The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required .

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

3 CAW-06-2095 (d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse .

It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system w include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors . It is, therefore, withheld from disclosure to protect the Westinghouse competitive position .

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2 .390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

CAW-06-2095 The proprietary information sought to be withheld in this submittal is that which is appropriately marked in NF-BEX-06-15 P-Attachment, " Westinghouse Input to Dresden Nuclear Power Station, Units 2 and 3 ; Quad Cities Nuclear Power Station, Units 1 and 2 -

Request for Additional Information Regarding Transition to Westinghouse SVEA-96 Optima2 Fuel" (Proprietary), for response to request for additional information , being transmitted by Exelon Nuclear letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk . The proprietary information as submitted by Westinghouse for the Dresden Units 2 and 3 and Quad Cities Units 1 and 2 is expected to be applicable for other licensee submittals in response to certain NRC requirements for justification of SVEA-96 Optima2 License Amendment Request.

This information is part of t which will enable Westinghouse to:

(a) Provide technical information in support of License Amendment Request.

(b) Assist customer to respond to NRC RAIs .

Further this information has substantial commercial value as follows:

(a) Westinghouse can use this information to further enhance their licen g position with their competitors.

(b) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse .

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses . Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended .

Further the deponent sayeth not.

Proprietary Information Notice Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval .

In order to conform to the requirements (A 10 CFR 2390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted) . The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2 .390(b)(1).

Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice . The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Westinghouse Non-Proprietary Class 3 NF-BEX-06-15 NP-Attachment Westinghouse Input to Dresden Nuclear Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units '1 and 2 -Request for Additional Information Regarding Transition to Westinghouse SVEA-96 Optimal Fuel January 25, 2006 ghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355 O 2006 Westinghouse Electric Company LLC All Rights Reserved 7016-NP.doc NF-BEX 06-15 NP-Attachment

NRC Request I The license amendment request was completed prior to the staff's approval of WCAP-15836-P-A and WCAP-15942-P-A. Now that these two topical reports have been completed, please update the applicability tables in Attachment 6 and the conditions and limitations tables in Attachment 7 to reflect the approved documents . Include the following :

a. Detailed descriptions of the plant-specific changes to the SVEA-96 Optimal fuel design and the evaluation to ensure mechanical compatibility with core components and co-resident fuel (WCAP-15942-P, Condition #2a) .

b. Detailed description of the control blade interference evaluation in accordance wit WCAP-15942-P, Condition #4.

Response

Updated Tables 16 and 18 of Attachment 7 (i .e., Reference 1), which correspond to the conditions and limitations of WCAP-15836-P and WCAP-15942-Pare attached . NRC formal approval of WCAP-15942-P has not been obtained . Therefore, conditions and limitations documented in Table 18 are based on the NRC's draft safety evaluation for WCAP-15942-Since WCAP-1 5836-P-A has not been issued, the applicability tables in Attachment 6 of Reference 1 remain valid .

The geometrical compatibility of SVEA-96 Optimal fuel with existing GNF (GE14) and FANP (ATRIUM-913 Offset) fuel, core internals and handling equipment in the Dresden Nuclear Power station (DNPS), Units 2 and 3 and Quad Cities Nuclear power Station (QCNPS), Units 1 and 2 plants has been evaluated according to References 2 and 3. The results from the geometrical study, based upon input data from Exelon Generating Company, LLC (EGC) and from Westinghouse experience show that the SVEA-96 Optimal fuel is compatible with existing fuel, core internals, fuel storage facilities and handling equipment during the design life of the fuel .

For detailed specifics refer to the resolution to Condition 2 in Table 18 attached .

The control rod insertability evaluation required by the draft SER for Reference 2 has been performed by Westinghouse for SVEA-96 Optimal fuel in the DNPS and QCNPS plants by combining plant specific assembly pitch and control rod dimensional information with the measured channel bow and channel creep experience database . The conclusion is that both the calculated maximum channel-to-control rod interference and available control rod insertion force-time for SVEA-96 Optimal in DNPS and QCNPS are bounded by proven Westinghouse successful operational experience and are demonstrated to be acceptable following the methodology defined by References 2 and 3 and Condition #4 of Reference 3. For detailed specifics refer to the resolution to Condition 4 in Table 18 attached .

7016-NP.doe Page 2 of 46 NF-BEX-06-15 NP-Attachment

Table 16 WCAP-1 5836-P Conditions and Limitations WCAP-15836-P Fuel Rod Design Methods for Boiling Water Reactors - Supplement I No . Condition I Limitation Resolution 1 STAV7.2 is approved for modeling BWR fuel with the The conditions are met via:

following limitations:

a. The pellet in SVEA-96 Optimal fuel used in

a. Solid U02 fuel pellet with a maximum gadolinia DNPS and QCNPS is solid U02 with the

]a,c content of maximum gadolinia content of

b. No substance beyond gadolinia and nominal trace b. No substance beyond gadolinia and nominal trace elements shall be added 5 be fuel pellet for the elements is added to the fuel pellet for the purposes of altering its physical characteristics. purposes of altering its physical characteristics

]a,c

c. Nominal fuel pellet density between percent c. Nominal fuel pellet density of [

theoretical . (between 92-97 percent theoretical) d- Fully RXA Zircaloy-2 fuel clad material . d. Fully RXA Zircaloy-2 fuel clad material

e. For fuel rods with clad liner (e .g . natural zirconium) e. Nominal liner thickness of mils the liner thickness shall be no greater than I .C

{nominal}. If . Peak rod average burnup limit of 62 GWd/MTU

f. Peak rod average bumup limit 62 GWdIMTU.

2 STAV7.2 shall not be used to model fuel above incipient The highest fuel temperature will be encountered in fuel melting temperatures . the fuel temperature calculation . The maximum fuel temperature is shown to be below the fuel melting temperatures .

3 STAY7.2 shall not be used to model fuel rods with an The maximum possible average cladding average cladding temperature above at temperature is from cladding strain or fuel any axial node . temperature anticipated operational occurrence (AOO) calculations, where the power is romped

]a " above thermal-mechanical operation limits (TMOL) . Even at the peak of the power ramps the average cladding temperature is below 4 STAY7.2 shall be used only within the range for which The SVEA06 Optimal fuel rod properties and fuel performance data were acceptable or for which assembly design are in the calibration and verifications discussed in WCAP-15836-P and verification database . The TMOL linear heat responses to RAls were performed. For example, generation rate (LGHR) which is the highest LHGR Section 3.8 describes a LHGR limit based upon the that can be experienced during normal operation is calibration and verification database of STAV7.2- lower than the LHGR limit specified in Section 3.8 of the safety evaluation .

5 Due to the empirical nature of the STAY7.2 calibration The released STAV7.2 is based on the approved and validation process, the specific values of the models- There is no update on the constants and equation constants and tuning parameters derived in tuning parameters .

WCAP-1 5836-P {as updated by RAls, e.g . Attachment 2 of Reference 0 become inherently part of the approved models. Thus these values may not be updated without Ifurther NRC review- Exceptions include the BWR (cladding corrosion constants (Table 2.2 .51), crud Ideposition constants (fable 2210, and rod nodal power Uncertainties for the BWR "Older" data (Uncontrolled and Controlled Cells in Table 3.3-1).

These exceptions will be addressed as part of the implementation methodology in WCAP-1 5942-P .

7016-NP.doc Page 3 J 46 NPBEMS15 NP-Attachment

Table 18 WCAP-1 5942-P Conditions and Limitations f ------I IWCAP-15942-P Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors- Supplement I to CENP-287 No . Condition I Limitation Resolution 1 Following the fuel assembly and fuel rod mechanical IThe amended methodology is followed in the design I

i design methodology described in WCAP-15942, as analysis for DNPS and QCNPS. Cycle specific amended by RAI responses, the licensee must ensure ( design changes and power histories will be checked that all of the design criteria are satisfied on a cycle- ;to evaluate whether this reference design analysis is specific basis . !still valid . If this analysis does not bound the specific cycle, a new design analysis will be performed, 2 The reference fuel assembly design SVEA-96 Optimal The calculations and evaluations performed for is approved up to a peak rod average burnup of DNPS and QCNPS are valid to a maximum fuel 62 GWd/MTU . assembly burnup of [ ]a,c, which supports a peak rod average burnup of 62 MWd/kgU .

(a) In order to ensure compatibility with DNPS and a . In addition to referencing this report in their LAR QCNPS Legacy fuel and core internals, the SVEA-96 submittal for implementing SVEA-96 Optimal, Optimal fuel (i .e ., fuel rods, active fuel length and licensees must include a description of the plant- ]a.c fuel channel) have been shortened by [

specific changes which are being made to ensure and "Style 2" in Section 5 of WCAP-1 5942-P was mechanical compatibility with core components and used, compared with previously evaluated and co-resident fuel . Further, the licensee must delivered SVEA-96 Optimal fuel to other reactors of demonstrate that these changes are within the the GE/KWU type .

envelope of approved plant-specific changes to the reference design description in Section 3 .1 Other plant specific changes are partial symmetrisation (same level as Legacy fuel) of the

b. Modifications to the fuel assembly design, beyond the originally asymmetric core lattice and minor mechanical compatibility changes identified in adaptations of the fuel assembly handle in order to Section 3 .1, will invalidate the staffs approval of the be compatible with Legacy fuel and core internals at SVEA-96 Optimal reference fuel design . The all conditions. Also the inlet piece bypass flow holes provisions described in Section 3 .1 .4 of are adapted so that the SVEA-96 Optimal fuel is WCAP-1 5942-P, "New Design Features", are not thermal hydraulically compatible with the Legacy fuel approved . and current core conditions .

These changes are consistent with RAI 7 of Reference 3 and are within the envelope of approved plant-specific changes to the reference design in Section 3 .1 of the NRC safety evaluation .

(b) There are no design feature changes to the Reference SVEA-96 Optimal fuel design defined in Chapter 2 of WCAP-15942 for DNPS and QCNPS other than those identified in the response to RAI 7 of Reference 3 .

3 The fuel mechanical design methodology and design The peak rod average burnup in this analysis is I criteria are approved up to a peak rod average burnup of :62 GWdIMTU . Additionally, 62 GWdIMTU . In addition :

a . The assembly design for DNPS and QCNPS is a . These methods are approved for application to the approved SVEA-96 Optimal design .

currently approved Westinghouse SVEA fuel assembly designs. b. Gap heat transfer calculations are the only analyses performed for non-Westinghouse fuel .

I b . These methods are also approved for the calculation of gap heat transfer coefficients (as described in Section 4 .4 and RAI#23) for mixed cores containing non-Westinghouse fuel designs.

7016-NP.doc Page 4 of 46 NF-BEX-06-15 NP-Attachment

I IWCAP-15942-P Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors- Supplement I ito CENP-287 No . Condition I Limitation Resolution 4 During initial implementation, licensees must submit to I The mechanical compatibility analyses to implement the NRC an evaluation of control blade interference ISVEA-96 Optimal fuel in DNPS and QCNPS confirm taking into account manufacturing tolerance, channel Ithat:

bulge, and channel bow over the life of the fuel assemblies (similar to RAI#15 response). Aspart ofthis a. The Maximum channel-to-control rod interference (blade and roller/pad) for DNPS and QCNPS is evaluation, the licensee must demonstrate the following:

less than that determined for la'C I l a. Calculated maximum channel-to-control rod interference (blade and roller/pad) must be less than b. The current data base used for the that determined for evaluation of control blade interference contains data from the 10x10 SVEA designs including the b, Westinghouse channel bow database remains valid. SVEA-96, SVEA-96+, SVEA-96 Optima and This demonstration must consider the materials and SVEA-96 Optimal designs. The mechanical manufacturing process employed in the fabrication of design of the channels for these designs has not the SVEA channels . been modified in a manner that would affect channel bow or bulge- The channel material has

c. Following the methodology outlined in RAI#1 5, the evolved from Zircaloy-4 in earlier designs to the calculated control rod force-time [ ((Paccumulator x current Zircaloy-2 channels . Furthermore, the Aannulus) tCR-73%) / MCR ] for the target plant must annealing process has been improved to provide be greater than or equal to me force-time parameter greater

]a,c . uniformity . Both of these changes tend to for [ reduce channel bow- Therefore, the entire data

d. Confirm SVEA channel experience is applicable for base provides a conservative description of the the specific application and continues to be bounded current SVEA-96 Optimal channels, with respect by the database presented in RAI#1 5 by assessing to channel bow, which will be installed in the the trend in control rod insertion time (e .g . the DNPS and QCNPS units.

number of "slovv" control rods) in US plants which c. The calculated control rod force-time for DNPS have implemented SVEA fuel channels since the time and QCNPS is greater than the force-time of issuance of this SER. This demonstration should parameter for identify the number of "slovv" control rods as well as the historical significance of these indications .

new Updates 9) me database reflecting channel bow data measurements may be used to address increasing trends in the numbers of slow rods . The will updated database be used as the bases to evaluate control blade interference

d. The ]a,c database in the response to RAI 15 of Reference 3 and used for the DNPS and QCNPS application is current. It will be updated as new data becomes available. Scram times in the DNPS and QCNPS units containing SVEA-96 Optimal fuel will be evaluated to detect any systematic increase in scram times or the numbers of slow rods which would indicate that the data base is not representative, 5 The lined SVEA fuel PCI threshold on LHGR must be For lined SVEA PCI thresholds that do not exceed shown to exceed the TMOL LHGR over the entire the TMOL, fuel PCI conditioning guidelines burnup range, otherwise fuel PC] conditioning guidelines applicable to non-lined fuel will be applied beginning applicable to non-lined fuel must be applied beginning at at LHGRs in excess of the lined fuel PCI threshold.

LHGRs in excess of the lined fuel PCI threshold.

I 7016-NP.doc Page 5 of 46 NF-BEX-06-15 NP-Attachment

action between the General Electric (GE) emergency core cooling system ante analyses of the GE14 fuel design and the Westinghouse ECI nafyses of the Optimal fuel design with respect to developing the bounding maxis average planar linear heat generation rate limits . Include within this descriptio plana he flow characteristics of each bundle design and how this information is ressed in ch respective ED The respo Question P-A describes the interaction between the ouse ECCS performance ce analyses performed by el present in the ing the transition to Westinghouse fuel.

n response to a loss of coolant accident (LOCH) for the sing three core models:

Uilib riled of 100% SVEA-96 Optimal f An equilibri riled of 100% GE1 4 fuel .

comprised of a mixture of SVEA-96 Optimal and performed using the 3D simulator, is used The thermal hydraulic c ibility analysis, which is ensure re model provides an accurate representation of

&on {e.g ., active core and the intra-assembly/

irite r-Z  ;; core pressure drop distribution, etc.) at nominal conditions .

The thermal ibility analysis, which is established based on extensive ical ulic information provided to Westinghouse by EGC, provides an accura anal hydraulic prediction of flow and pressure distributions for a variety of core configurations . The following table summarizes the important thermal hydraulic features of the GE14 and SVEA56 Optimal fuel designs for a full core of the designated fuel assembly in the u a ;c 7016-NP.doc Page 6 Y 46 NPBEkOT1 5 NP-Attachment

istics are included in the LOCA model, which is tuned to match the

)onding pressure drops, flow splits, etc. After ensuring that the three LOCA core mode

~urate at nominal conditions relative to the thermal hydraulic compatibility analysis, a esponse analysis is performed to determine the system response for each The system response model also includes boundary conditions to the hot channel model, which to determine the response of the hot assembly. The boundary conditions from the hot assembly are used to determine the thermal response of the fuel rods (i.e ., peak cladding erature, maximum cladding oxidation) and ultimately to develop the (maximum linear heat generation rate MAPLH, ) limits . The impact of different core configurations boundary conditions i of three key events, which impact the MAPLHGR analysis . These are [

qC I

consider the system response from the three configurations to determine the one to evaluate the Optimal MAPHGR limit . For determining the limiting system response, Westinghouse will evaluate the time of uncovery, the time Core Spray pumps achieve rated flow, and time two-phase conditions are re-established . Should the system response for the mixed core be more limiting than that for GE1 4, Westinghouse will inform EGC to have GNF aluation of the impact of the mixed core on the GE1 4 MAPLHGR limits for the transition to SVEA-96 Optimal fuel.

7016-NP.doc Page 7 N 46 NF-BENOT15 NP-Attachment

NRC Request 5 Discuss the applicability of seismic/loss-of-coolant accident methodology in CEN PD-288-P-A to the SVEA-96 Optimal fuel design . Include a discussion of the mechanical testing done on the Optimal grids .

Response

CENPD-288-P-A, Reference 4, describes the general Westinghouse methodology which demonstrates that the Westinghouse reload fuel assembly satisfies the following design bases under a postulated seismic/LOCA event:

a. Fuel fragmentation will not occur as a result of combined normal operation, seismic, and LOCA loads.

Control fired.

b. rod insertability will not be

c. Spacer grid distortion will not be sufficiently great that fuel rod coolability would be prevented .

Section Or 2 of Reference 4, the seismic/LOCA evaluation is performed for each plant application of Westinghouse BAR fuel. The methodology is defined in a clear and generalized format that can be applied :

To both Westinghouse and non-Westinghouse des ned BWR fuel.

all In BAR reactors (e .g . E0WR/2 through EMR16) .

0 Accommodating a variety of plant licensing bases and available seismic and LOCA data .

Reference 4, has been approved by the NRC with the conclusion that it presents an adequate and acceptable methodology to evaluate all Westinghouse BWR fuel assemblies subjected to postulated seismic/LOCA events with no restrictions imposed . Therefore, it can be concluded that CENPD-288-P-A is applicable to the Westinghouse SVEA-96 Optimal fuel assembly.

Reference 4, documents mechanical tests of spacer grids that have been performed to verify the performance under seismic-type loads . The primary tests performed to address potential seismic loads are the lateral load cycling tests . [

kc I

The SVEAQ6 Optimal fuel is a further development of the SVEA-96 design . The SVEA-96 Optimal spacer grid design is based on the same Westinghouse SVEA-96 grids with the same 7016-NP.doe Page 8 of 46 NPBEkOV15 NP-Attachment

principal design of the grid cell and of the same material . However, there are differences between the two fuel types that may lead to different dynamic responses under a seismic load.

tc I I

]"C For example, for a typical application these design changes were determined to change the a

natural frequency of the SVEA-96 Optimal assembly by about [ ] '` percent relative to SVEA-96. This leads to less than 1 percent change in deflection.

As discussed in Section 8.3 of Reference 2, Westinghouse has also performed lateral load cycling tests with lowcycle fatigue for the SVEA-96 Optimal fuel to qualify spacer and channel

]JC .

welds for seismic loads. The test conditions were [

These tests were performed at room temperature, and scaling factors were used to translate test results to operating conditions in accordance with ASME Section III, Appendix 11-1520. The scaling factors include the effects of the temperature and irradiation as well as experimental uncertainty.

The tests have verified that the spacer grids and welds will withstand the following lateral seismic type acceleration at operating conditions without failure and with negligible deformation:

phc 0 Spacer grid: I 0 Channel welds : [

For more detail refer to Section 8.3 of Reference 2.

For IDNPS and QCNPS, the mechanical behavior of the SVEA-96 Optimal fuel during a postulated combined Safe Shutdown Earthquake (SSE) and Loss of Coolant Accident (LOCA) event is currently in progress and will be completed prior to plant start-up . The methodology for the calculation of stress intensities and component deflections documented in Reference 4 will be followed . The structural analysis of the fuel assembly is based on fuel support and core grid response spectra for SSE and channel pressure load from the most limiting LOCA event. The acceptability of the results will be evaluated against a set of material and component acceptance criteria or experimentally based acceptable external forces, Reference 4 and 5, consistent with the USNRC Standard Review Plan, Section 4.2, Reference 8, and ASME Section 111, Appendix F, Reference 9. All tests necessary to support the methodology have been performed or are judged to be unnecessary.

7016-NP.doc Page 9 of 46 NF-BEX-06-15 NP-Attachment

NRC Request 7 Section 2.3 of the license amendment request identified a change to the Westinghouse EGGS n methodology for the transition to SVEA-96 Optimal .

a . Per 10CFR40.56, EGG needs to submit for staff review :

Justification that the Westinghouse EGGS Models are acceptable for and properly applied to Dresden and Quad Cities .

ii . Results of the plant-specific EGGS evaluation (detail sufficient for staff review) .

Response

A report will be provided upon completion to justify the acceptability of the application of the Westinghouse ECCS evaluation methodology for the transition to SVEA-96 Optimal fuel at DNPS and QCNPS. The report will describe a single `Unit 5' model that bounds, from a LOCA perspective, all four DNPS and QCNPS units. The report will describe the application of Westinghouse methodology in sufficient detail to demonstrate that the EGGS models are applied properly and in conformance with all limitations / conditions placed on approved topical re ports.

This report will provide the basis for future 10 CFR 50.46 evaluations of plant changes; errors discovered in the approved evaluation model ; or errors in the application of the approved evaluation model.

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NRC Request 8 Section 4.3.1 states, "Since the raw CPR data that was used to develop the legacy fuel vendor's CPR correlation will not be provided, a conservative adder will be applied to the legacy fuel operating limit minimum CPR which satisfies the 9555 statistical criterion." Demonstrate that the adder meets the 95/95 criterion .

Response

7016-NP.doc Page 11 of 46 NF-BENOT15 NP-Attachment

GEC 4 legacy fuel. The tc I

"c USAG14 =CPR correlation for GE14 CPR correlation for SV f = massflux 00I M 2 _ S )

I bly exit qxressure (bar) h = assembly inlet enthalpy {J/gm) ction coefficie 7016-NP,doe Page 12 of 46 NPBET0415 NP-Attachment

3,ENRdc P~ ]a a 4 NRB~x~ ,SNRk!!rn

]a,c Table 8-1 Cosine Axial Power Shape (node 1 = bottom}

0 .349 0 .496 a 636 a769 a893 06 1106 7 1 .107 8 1 .194 9 1 .267 10 1325 11 1366 12 1392 1 .4 14 1 .392 15 1 .366 16 1 .325 17 1 .267 18 1194 19 1 .107 20 1 .006 21 0.893 22 0 .769 0 .636 24 0 .496 25 0 .349 r,c 701WNP .doc Page 1 4 of 46 NF-BEX-06-15 NP-Attachment

Example : Comparison of CPR_Ex stinghouse far GE'14 Legacy fu tiplier C (note that this is labeled as an adder to the it will in effect increase the OLMCPR) to the OLMCPR calculation PC 7016-NP-doc Page 1 5 of 46 NRBENOT15 NP-Attachment

In Attachment 6, page 5 of 11, the last paragraph alludes to the Westinghouse Topical Report WCAP-15942-P as containing the Westinghouse experience base . Please provide this data base in Tabulated form, including as much detail as possible regarding Extended Power Uprates (EPU) and operation with high exit void fractions. That is specifically :

a . Demonstrates quantitatively and qualitatively, that the Lattice/Depletion code systems, and that the current uncertainties and biases established in the Lattice/Depletion code systems remain valid for the neutronic and thermal-hydraulic conditions predicted for the EPU operation. Specifically, demonstrate the uncertainties and biases that are used in the licensee's reactivity coefficients (e .g . void coefficient) are applicable or remain valid for the neutronic and thermal-hydraulic conditions expected for EPU operation.

b. Demonstrate quantitatively and qualitatively, that the fuel isotopic validations and testing performed in the Lattice/Depletion code systems remain applicable for prolonged operation under high void conditions for the fuel lattice designs that would be used for the expected EPU core designs.

c. Demonstrate qualitatively and quantitatively that the Westinghouse neutronic methodology experience base and demonstrate that the Westinghouse methodology is applicable to EPU conditions, specifically to EPU conditions at Dresden Nuclear Power Station (DNPS) and Mad Cities Nuclear Power Station (QCNPS) .

d. Provide any validation data in support of the Westinghouse neutronic methodology prediction capability by comparison to gamma scans and Transverse Incore Probe (TIP) core follow benchmarking based on the current fuel designs operated under the current operating strategies and core conditions . This request pertains to any recent fuel, such as the SVEA-96+ and OPTIMA-2, in particular for first cycle and second cycle fuel.

Response

In the Westinghouse BVVR methodology, the Lattice/ Depletion code system is used to generate cross sections and other cell data for the core simulator. Uncertainties, biases or even reactivity coefficients are neither generated by nor computed directly from the Lattice/Depletion code system. The cross sections and cell data are generated at the particular plant's conditions, yet as shown in Table 9.1, the DNPS and QCNPS extended power uprate (EPU) conditions fall within Westinghouse's experience base. The plants in which Westinghouse BWR fuel has been used am referred to in bold type in Table 9.1 . The application to the DNPS and QCNPS units is indicated by the use of italics for these plants .

The Westinghouse BWR methodology uses uncertainties associated with the power calculations performed by both the Lattice/Depletion code system and the 3D core simulator. The nodal, assembly and pin nodal relative power uncertainties currently used by Westinghouse and included in Westinghouse Topical Report CENPD-390-P-A were noted in Attachment 6, page 9, first row of Table P-1 . Those uncertainties were generated from comparisons against 7016-NP .doc Page 1 6 of 46 NR BEYOM 5 NP-Attachment

measurements for four plants (see page 90 of CENPD-390-P-A) . Since approval of CENPD-390-P-A, Westinghouse has performed additional gamma scans, to support the introduction of new fuel types (including SVEA-96 Optimal}, as well as reactor thermal power (including EPUs), well beyond the power level and bundle average power level at DNPS and QCNPS. Those additional gamma scans were presented in Attachment 6, page 9, rows two through four of Table P-1 . As can be seen in the table, neither the introduction of new fuel types, nor higher power levels have degraded the accuracy initially documented in CENPD-390-P-A .

Westinghouse Topical Report CENPD-390-P-A Chapter 3 presents the qualification of the Lattice/Depletion code system and its associated library. In that chapter, several critical experiments are modeled with the Lattice/Depletion code system. Nevertheless, the chapter starts with the following statement - "The primary application of PHOENIX is to generate the few group nodal cross sections and other physics constants for POLCA . Therefore, the benchmarking of POLCA to plant data described in Chapter 5 provides the best overall qualification of PHOENIX." Thus, isotopic validation is not performed directly with the Lattice/ Depletion code system, but more as part of an integral method, including the core simulator. Nevertheless, Westinghouse continually evaluates its Lattice/Depletion code system by comparing calculated global parameters (reactivity, power distributions, fission/capture rates) against higher order methods.

Westinghouse Topical Report CENPD-390-P-A Chapter 5 presents the qualification of the 3D core simulator. In that chapter, multiple comparisons are presented, including gamma scans and traversing in-core probe (TIP) instrumentation comparisons for four different plants . The gamma scans and TIP comparisons included in CENPD-390-P-A include SVEA-96 fuel assemblies . As previously mentioned, additional gamma scans have been performed to address newfuel types and more demanding operating conditions . Regarding new TIP comparisons including SVEA-96 Optimal at more demanding operating conditions, Figure P-2 in page 10 of Attachment 6 notes very consistent results for multiple cycles at [

]". The figure notes the nodal and radial root mean square (RIVIS) differences, as well as the fraction of loaded fuel containing part-length rods . This figure notes that for more challenging conditions than those at DNPS and QCNPS, the introduction of SVEA-96 Optimal did not cause a degradation in the TIP comparison results with the Westinghouse neutronic methods .

Westinghouse has previously applied the Lattice/ Depletion code system as well as the core simulator for neutronic and thermal-hydraulic conditions that cover the EPU conditions at DNPS and QCNPS. In addition, Westinghouse has continued to validate its power uncertainties with additional testing and measurements at more demanding conditions than those at DNPS and QCNPS. Westinghouse also continually evaluates the Lattice/Depletion code system and core simulator performance, and how they are used within the BWR methodology. It is Westinghouse's conclusion that its neutronic methods are capable of accurately modeling the EPU conditions at DNPS and QCNPS.

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I - Experience Data Base

• Plants in which Westinghouse BWR fuel has been used .

7016-NP .doc Page 18 of 46 NPBEkO& 1 5 NP-Attachment

NRC Request 10 In Attachment 6, page 6 of 11, the discusses briefly the contents of CENPD-390-P-A.

a . Does this topical include OPTIMA-2 data/analyses?

b. Does this topical contain TIP pin power comparisons for normal and extended power operations?

Response

The comparisons presented in Westinghouse Topical Report CEN PD-390-P-A Chapter 5, "POLCAQualffication," predate the introduction of the SVEA-96 Optimal fuel. Nevertheless, the information provided in Section 4.0 of CENPD-390-P-A, "POLCA," does include model descriptions for part-length rods assemblies . One of the main objectives of this new version of POLCAwas the treatment of part-length rods assemblies .

Westinghouse Topical Report CENPD-390-P-A includes TIP analyses for four different plants

{see Table 5.6 on page 9Q. The information in the Topical Report includes both normal and EPLJ conditions for the reactors identified as A, B, and C. The information for Reactor D is for normal operating conditions only. The topical report includes pin power comparisons for two different plants (see Section 5.3.2 on page 72) . The assemblies analyzed were for reactors prior to undergoing their EPUs. However, those reactors at pre-uprate conditions were at a higher power density than DNPS and QCNPS.

Although Westinghouse Topical Report CENPD-390-P-A does not include results with SVEA-96 el nor with the challenging EPU conditions present today, Westinghouse has performed additional gamma-scan and TIP comparisons to address new fuel types as well as increased power levels and more challenging operating conditions . The nodal, assembly and pin nodal relative power uncertainties currently used by Westinghouse and included in Westinghouse Topical Report CEN PD-390-P-A were noted in Attachment 6, page 9, first row of Table P-1 . Rows two through four present the results for the additional gamma scans. As can be seen in the table, neither the introduction of new fuel types, nor higher power levels have degraded the uncertainties initially documented in CENPD-390-P-A . Regarding newTIP comparisons including SVEA-96 Optimal at more demanding operating conditions, Figure P-2 in page 10 of Attachment 6 notes very consistent results for multiple cycles at [

]a,c . The figure notes the nodal and radial RMS differences, as well as the fraction of loaded fuel containing part-length rods . This figure notes that for more challenging conditions than those at DNPS and QCNPS, the introduction of SVEA-96 Optimal no (muse a degradation in the TIP comparison results with the Westinghouse neutronic methods.

7016-NP.doc Page 1 9 of 46 NF-BEX-06-15 NP-Attachment

NRC Request 1 1 Provide the TIP and Gamma comparisons and PROTEUS results, discussed in the 2 "d , 3 and 4"'

Paragraphs on page 6 of 11, Attachment 6.

Response

The second and third paragraphs on page 6 of 11, Attachment 6 of Reference 1, provide background information on Westinghouse's methods for establishing the power uncertainties, which includes plant TIP comparisons and pool-side gamma scan measurements . The fourth paragraph refers to experiments performed at the KRITZ facility and at the I_WR-PROTEUS facility. Attachment 6 already contains results for the latest set of TIP comparisons, gamma-scan measurements, and the PROTELIS experiments.

Westinghouse Topical Report CENPID-390-P-A presents results for the KRITZ facility experiments, as well as several plant TIP comparisons and pool-side gamma scan results. TIP comparison results are presented in Chapter 5 for four different plants . Gamma scan results are also presented in Chapter 5, for two sets of measurements . Since the approval of the Topical Report, Westinghouse has performed additional TIP comparisons and gamma-scan measurements to address new fuel types as well as increased power levels and more challenging operating conditions . The nodal, assembly and pin nodal relative power uncertainties currently used by Westinghouse and included in Westinghouse Topical Report CENPD-390-P-A more noted in Attachment 6, page 9, first row of Table Poll . Rows two through four present the results for three additional sets of gamma scan measurements . Regarding new TIP comparison results, Figure P-2 in page 10 of Attachment 6 presents results for multiple cycles at [ ]" . The figure notes the nodal and radial RMS differences, as well as the fraction of loaded fuel containing part-length rods .

The PROTEUS experiment results are also included in Attachment 6 . Rows five and six of Table P-1 and Figure P-1 present those results. The second paragraph on Attachment 6, page I of 11 provides some discussion on the PROTEUS experiment results .

NRC Request 12 In Attachment 6, page 7 of 11, the first four paragraphs on this page, and the Tables that go with them, require further clarification.

Response

lotion and equations used for the statistical calculations are included on pages 73-74 of Westinghouse Topical Report CENPD-390-P-A. In the Topical Report, the differences between measurements and POLCAcakulated values are noted as RIVISoverall, RIVISradial, and RMS,,,,i,,. The "overall" label represents the nodal differences, whereas the "radial" label represents the assembly differences. Note also that in the Topical Report, the differences are left in percentages.

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On page 122 of Westinghouse Topical Report CENPD-390-P-A, the relative fuel rod power uncertainty of [ ]", relative nodal power uncertainty of [ ]", and relative assembly power uncertainty of [ ]a,c are noted . Those same values are noted as "fractional standard deviations" in Attachment 6, page 9, first row of Table P-1 . Thus, the term fractional standard deviation implies the uncertainty, in fractional (not percentage) form .

NRC Request 13 In Attachment 6, page 8 of 11, the t paragraph alludes to pin power testing with results obtained for the mid-planes .

a. Does Westinghouse have any plane pin power behavior, particularly at very high exit void fractions?

Provide qualitative description of the void data base and the associated correlation .

Specifically describe the uncertainty associated with the data gathering, specifying the uncertainties currently applied to the void fraction correlation and justify its applicability for EPU conditions .

Response

The pin power testing performed in the PROTEUS facility is performed at the mid-plane. The reason is that the facility is a small critical core, with significant axial leakage . In order to facilitate the validation of the lattice codes, the measurements are performed at the mid-plane, where the spectral conditions are least sensitive to leakage . The experimental conditions in the irection are constant for measurements performed at [ ]", the entire experiment's axial distribution has constant (non-voided) density. Similarly, for measurements performed at ]ax, the entire experiment's axial distribution is set to

]a,c.

I As noted in Attachment 6 of Reference 1, page 10 of 11, Figure P-1, the PROTEUS experiments were performed at four different conditions . The first set, [

qC I

In connection with the introduction of 10x10 fuel designs with part-length rods, Westinghouse performed new void measurements at its FRIGG loop to confirm the validity of the void correlation . The new measurements were performed on a SVEA-96 Optima model, with

]". Figure 13 .1 shows the void correlation prediction - measured void results, as a function of measured void . Two things to note in the figure are the lack of a [

PI C .

7N&NR&c Page 21 of 46 NF&EXTU 5 NP-Attachment

asurements taken cover the range observed at DNPS and PU conditions, Figure 112 was generated. This figure presents the a n for three sample hot channels at [

]a,c. Based on the values shown in Figure 13 .2, the void rformed at FRIGG clearly cover the range observed at DNPS and QCN PS Although there is uncertainty in the measurements and data g Figure 13.1 urements Results xC 7&TNRWc Page 22 of 46 NF-BEX-06-15 NP-Attachment

Figure 13.2 DNPS an annel "C 7016-NP .doc Page 23 of 46 NkBETOW5 NP-Attachment

NRC Request 14 In Attachment 7, page 9 of 43, the justification provided on the next three pages to extend the AA78 slip correlation to pressures beyond those reviewed and approved in the topical report, equire additional quantitative technical justification. For example, nothing was stated regarding the possible effects on the uncertainties introduced due to extrapolation of the Westinghouse void correlation beyond its current data base . Please provide qua iption of the void data base and the associated correlation. Specifically describe the uncertainty associated with the data gathering, specifying the uncertainties currently applied to the void fraction correlation and justify its applicability for EPU conditions .

Response

Description of the AA78 void correlation data base The AA78 slip correlation is described in the BISON Topical Report RPA 90-90-P-A. This correlation is basically a bubble flow correlation modified to cover annular flow for BWR fuel bundle . [

"C I

fit The correlation is a best to void measurements performed with full-scale (36 and 64 rods) test sections in Westinghouse's FRIGG test loop. The original recommended range of applicability was:

Pressure : 3 .0 to 9.0 MPa (435 to 1305 psia)

Max flux : 500 to 2900 kg/M2 S (0.30 to 2 .1 Mlb/h_ft2)

Quali 0 to 1 .0 Covered ranges t arty FRIGG void measurements were:

Table 14-1 Covered Ranges in the AA78 Database Mass Flux Steam Quality Void Fraction 2S)

Test Section Pressure (bar) (kg,M max} max)

OF-36 30-90 550-2900 40 90 OF-64A 48,68 500-2500 40 90 OF-64B 68 500-2000 55 95 Additional void measurements were later performed for SVEA-96 geometries (sub-bundle test sections) which extended the validity of AA78 correlation to 400 kg/M2S - The void predicted by the AA78 correlation was compared to these new measurements and extrapolation below the data range for mass flux is considered acceptable at least down to 400 kg/ mss.

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This new data coven; following Table 14-2 Add data utilized for V&V of the AA78 void correlation Mass Flux Steam Quality Void Fraction Test Section Pressure (bar) ftlm 2 s} max) max)

SF24VA 55,70 400-2000 35 90 SF24VB 55, 70; 80 400-1625 40 87 The error distribution and standard dev id correlation as a function of the void is showed in Table 14-3 and the comparison against each measurement series in Table 104 .

Table 14-3 Error distributio function of the AA78 predicted void Table 14-4 Mean error and standard deviati redicted void compared to the measured void for the different series I void correlation correlation is based on a larger data base which includes not only rod bundle so measurements from heated rectangular channels and round tub tiara of the correlation is given in EPRI Report NP-2246-SR, "A Mechanistic Model for Predicting Two-Phase Wd Fraction for Water in Vertical Tubes, Channels, and Rod Bundles,"

G.S. Lellouche and B.A. Zolotar, 1982.

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for The statistical Analysis of the Model versus Data the different type of measurements is provided in Table 3, 8, and 11 of the EPRI report and summarized in the following table. In addition EPRI NP-2246-SR Table 13 gives the Model versus Data - Pressure and Flow Range Comparison .

Table 14-5 Mean error and RMS error of the EPRI predicted void compared to the measured void for the different type of experimental data 01010 0.022 440 Information provided during the NRC revision of be PA78 void correlation (Topical Reports RPA-90-90-P-A and CEN PD-292-P-A)

During the NRC review of the Topical Report RPA 90-90-P-A, questions regarding the void models were discussed further. Some of the information provided in responses is relevant to

]a,c .

the discussion of the applicability of the correlation to pressures higher than [

Question 5 regarding the limitations of several correlations, including AA78, are answered on RPA-90-90-P-A pages Q5-1 to Q5-6 and included comparisons with FRIGG loop data . The following text has been extracted from the response regarding the AA78 void correlation.

"The verified data range covers most BWR applications . However, in some extreme cases, such as design basis pressurization transients (MSIV closure without position scram) or trip of all recirculation pumps, the limits of the above data range may be exceeded . However, the dependencies in pressure and mass flux are smooth and continuous, and the correlation prediction outside the above range follows the expected trend."

To justify that extrapolation beyond the test conditions is acceptable, two figures, 05.1 and Q5 .2, were provided . Figure Q5.1 plots measured void against steam quality for two pressures, 7 and 9 MPa (1015 and 1305 Asia}, at the same inlet subcooling . Also shown are the BISON calculated curves for various pressures. These calculated curves show that there is a smooth trend in void as a function of pressure . Figure Q5.2 shows measured versus calculated void at different pressures, and demonstrate that there is no significant trend in the error as a function of pressure .

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Question 024 requested further justification of he use of he void and boiling correlations in

]a,c BISON at pressures higher that [

Comparison with other correlations with somewhat larger ranges of applicability has verified that the correlation behavior is also correct outside the above ranges . Further discussion and justification is provided through the response to NRC Question 24 on RPA-90-90-P-A pages to Q24-6. Comparative graphs, of the same type as now provided in Reference 1, of pressure trends up to [ ]" . The graphs combined with the Solberg boiling/

compare void change trends predicted with AA78 with he Le Ilouche-Zo Iota r EPRI slip correlations described above .

condensation model and The EPRI correlation has been verified for a wide range of pressures . It was developed to fit not only the rod data which forms the basis of he AA78 correlation, but also other data including measurement in rectangular channel experiments at 10 .3 and 11 .0 MPa (1493 and 1598 psia).

Thus, it serves as a reference for the variation of void fraction with pressure for a range of geometri The following text has been extracted from the response to NRC Question 24:

te I

Comparison of this figure with the corresponding curves calculated with AA78 and the Solberg models using parameters derived for a single channel application (AA, Figure Q24.2), and using parameters for application to core average conditions (W, Figure 024.3), and also with curves calculated using the modified Bryce-Holmes correlation (Figure Q24 .4), indicates that the

]a,c change of void fraction with pressure over the range is the same for all methods. "

The application of he P178 void correlation to pressures up to was justified through the response to Questions Q5 and Q24.

The matter was further discussed in the supplement to the Topical Report RPA-90-90-P-A, CENPD-292-P-A "BISON - One Dimensional Dynamic Code for Boiling Water Reactors :

Supplement 1 to Code Description and Qualification," July 1996. This supplement to the BISON topical report was submitted, among other improvements, to change the boiling and condensation model (core void profile) [

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The qualification was provided in Section 6 .5.3.2 (comparison against the Peach Bottom Turbine Trip data) and in Appendix A in the response to NRC Question Al to CENPD-292-P-A.

The same qualification as the one performed in response to Question Q5 to RPA-90-90-P-A, was repeated for the EPRI boilingloondensation model in combination with the AA78 slip (void) correlation . The results of the prediction against the FRIGG loop data are presented in Figures A1-1 and A1-2. These figures show that the correlation gives comparable results with no

]a,c .

systematic deviations over the entire range of void fractions up to [

Justification for Extent, the Validity Ran 78 Correlation To calculate the pressure response during an sient without scram (AT\/\/S) up to the acceptance criterion of 1500 Asia, [

]' , c This range increase is supported by extended comparative graphs of the same type as the ones presented in the response to NRC Question 24 to RPA90-90-P-A shown below.

The two differential voids versus steam quality figures for AA78 and EPRI respectively, show that both correlations have the same trends . [

]Mc The AA78 correlation is as shown above verified against measured data for pressures up to

[ ]". In Figure 6.4 of Topical Report CENPD-292-P-A, "BISON - One Dimensional Dynamic Analysis Code for Boiling Water Reactors: Supplement 1 to Code Description and Qualification," the RMS error of the AA78 correlation as implemented in BISON is given to be I Y" by direct comparisons to measurement data . The mean error is When extrapolating further a comparison with the EPRI correlation is used.

The EPRI void correlation (equivalent to the Chexal-Lellouche drift flux correlation) is described in the Paul Coddhgton and Rafael Macian paper "A study of the performance of void fraction correlations used in the context of drift-flux two-phase flow models," Nuclear Science ring and Design, 215 (2002) 199-216 . In this paper, void fraction results were compared to a wide-range of experimental data with various geometry, inlet subcooling, power distribution, and pressure values (up to 15 MPa = 2176 Asia} .

Comparing the differential void changes versus 7.0 MPa and calculating the standard deviation and the bias between AA78 and EPRI for pressures between generates the following graphs .

7016-NP.doe Page 28 of 46 NF-BEX-06-15 NP-Attachment

Figure 14-1 [

X0 lanabon far [ ]~ C when extrapolating and the good between EPR I and AA78 at pressure higher than the AA78 data base maximum pressure (FRIGG loop measurements) is given by the fa the lion of comparison against experimental data provided in the response to and Q24 to RPA-90-90-P-A demonstrate that there is a smooth trend in void as a function of pressure and that there is no significant trend in the error as a fu pressure . [

cc I

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d by comparing the AA78 void to other void correla experimental data for a wider ran of pressures, similar to the EPRI void correlation which to 11 Also Table 13 of the EPRI report NP-2246-SR shows the bias with pressure . The comparison between AA78 void correlation to other methods as shown in Figures 024 .1 to Q24.4 and the figures in of Reference I (also presented below) in e predicted over the range [ I" is the same for EPU conditions with increased flow window, the core average void is expected to increase since the core average power 4 higher even though the sub-cooling also increases due to edwater flow. However, the highest void fractions occur in the hot channels. The ions still have approximately the same exit void fraction, since they still are limited by the thermal limits (e .g. CPR, that li die power) . At EPU condition the highest power channels have practically unchanged exi coons. The mai at EPU conditions is that more channels have higher powers .

For this reason, all correlations valid at high voids {e.g . AA78 which is based on rod bundles void measurements up to [ ]a ") are still within range at EPU conditions . Further justification of the applicability of the void correlation to EPU conditions and the comparison of the POLCA predicted void to the more recent FRIGG measurement for SVEA-96 Optimal is provided in the response to NRC Request 13 above .

Figure 14-2 [

7016-NP.doc Page 30 of 46 NF-BEX-06-15 NP-Attachment

Figure 14-3 [

a,c 7016-NP.doc Page 31 of 46 NRBEkOT1 5 NP-Attachment

oiling cr or specification that applies to the TIP and the local power PD-300-P-A, the Westinghouse BWR fuel assembly flow within the same range as the original plan Trent resident fuel. In addition, the BWR fuel assembly re sufficient flow to the water cross, in order to prevent significant he water cross at full power. An axially-averaged void co inlets to fuel assembly bypass flow rate to meet the design criteria . Therefore, inghouse Optimal fuel in the DNPS and QCNPS n the existing voiding criteria or specification or the pe TIP and the local power range monitor {t_PRnnj reap ding in the bypass regions on the accuracy of the simulated in-core detector 0 addressed here . An important result of the POLCA calculation for use in comparing LCA predictions to measurements is the simulation of the signals from the neutron and gamma sensitive detectors in the core . The simulated detector signals are determined for both TIP and the LPRM . The neutron sensitive response calculation is based on computing the n rate induced in the detector by the fast and thermal flux at the detector region .

odel (explicit detector modeling during lattice calculation, together with a detector core simulator) is used to simulate the response of neutron-sensitive The detector reaction rate is calculated via :

a.,c In the core simulator, [

The detector formula used in the simulator retie known homogeneous flux .

VIC 7016-NP.doc Page 32 of 46 NPBETOW5 NP-Attachment

XG a .C The The best evid( reliability of the model is the excellent agreement obtained between simultaneous ent of TIP distributions and bundle gamma scan. For details refer to response to N 10.

Evaluate capability of the licensing code systems, including the core simulator, i determining tial for bypass voiding .

in the response to NRC Request 15, the Westinghouse BWR fuel assembly is maintain the inter-assembly bypass flow within the same range as the original plant n or within the same range provided by the current resident fuel. In addition, the assembly is also designed to assure sufficient flow to the water cross, in order to prevent nificant boiling in the water cross at full power. An axial averaged void content of [

]'" . The inlets ly bypass flow regions sized to required flow rate to meet the design criteria .

The system simulation code, POLCA, is used to calculate that bypass flow rate in the bypass region . The size of bypass region is used by POLCA to determine the bypa flow rates utilizing the same ther I hydraulic governing equations of conservation of mass, momentum and energy, which ar d in the active region of the fuel assembly. Similarly, the 7016-NP.doe Page 33 of 46 NPBENOT15 NP-Attachment

constitutive relations that close the solution of the mentioned governing equations in the active region are also used in the bypass region . Since the active region and the bypass region communicate with each other through several paths, the equations are solved iteratively until the calculated pressure drop through the active and bypass regions are reasonably close . [

qC I

The Westinghouse safety analysis and design codes, BISON and GOBLIN, model the transport of momentum, mass and energy of single phase and two phase coolant in the core and bypass channels and the external coolant loops. No distinction is made between the active and the bypass regions of the core . The same conservation equations and constitutive relations are used in the core and bypass regions . The conservation equations are solved iteratively until the pressure drops in the active and the bypass regions are reasonably close.

The process described above is consistent with other thermal hydraulic codes used by the industry, such as NAPRE, RETRAN, REAP etc, with no restriction on the amount of bypass flow through the bypass channels .

NRC Request 17 Provide evaluation and discussion of the lattice/depletion code capability to generate the cross-section with voiding in the in-channel water rods and bypass.

Response

The PHOENIX two-dimensional physics lattice code is used to generate cross sections used by the core simulator code POLCA, including the detector relative signal, as a function of I

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Evaluate EPU core neutronic an ulic conditions and state for EPU core designs operating c( ate or transient events .

cider operat is the [

r,c r ardless of EPU. DNPS and sled in the response to NR in Table 9-1, even after EPU ies, primarily, lead to [

I kc nse to l,JRC Requests 15 through 17, there is the possibility of a small hannels during normal operation and anticipated operational

, discussed that small amounts of boiling in the bypass region have I a,c Figure 18-1 Average Void Con t Before/After EPU a_--o 7016-NP.doc Page 35 of 46 NPBEVO&15 NP-Attachment

i be seen, followi uprate results in a negligible change in the car e 18-2 Void Content Before/After EPU wo 7016-NP.doc Page 36 of 46 NPBEkO&15 NP-Attachment

NRC Request 19 In August 30, 2004, General Electric Nuclear Energy (GENE) issued a Part 21 report (ML042720293), stating that using limiting control rod blade patterns developed for less than rated flow at rated power conditions could sometimes yield more limiting bundle-by-bundle MCPR distributions and/or more limiting bundle axial power shapes than using limiting control rod patterns developed for rated flow/rated power in the SLIVICPR calculation. GNF-A evaluated the plants operating at the MELLLA operating domain and concluded that the potential exists for conditions more limiting SLMCPR at the nonrated low for plants currently operating at the MELLILA domain as well . GNF-A also evaluated the plants operating at the MELLLA operating domain and identified four plants that may have more limiting SLIVICPR calculated at the plants submitted amendment requests increasing ow statepoint . The affected their SLMCPR value. The staff understand that Framatome did not issue a Part 21 reporting on the SLIVICPR methodology that addresses the calculation of the SLMCPR at minimum core flow and offrated conditions similar to GENE's Part 21 report (ML042720293) . The following topics pertain to Framatome's methodology for calculating the SLIVICPR at minimum core flow at rated power statepoint .

a . Provide reference(s) to the applicable sections of the SLMCPR Westinghouse methodology that specifies the requirement to calculate the SLMCPR at the worst-case conditions for minimum core flow conditions for rated power. Please demonstrate to the staff that the SLIVICPR is calculated at different statepoints of the licensed operating domain, including the minimum core flow statepoint and that the calculation is performed for different exposure points .

b . Discuss or reference the applicable Sections/Chapters that addresses what rod patterns are assumed in performing the nonrated flow SLMCPR calculations . State how it is established that the rod patterns assumed in the SLIVICPR calculations for rated power, flow, and minimum core flow conditions, would reasonably bound the planned rod pattern that DNPS and QCNPS; would operate under EPU conditions .

c. For implementation of ARTS/MELLLA using Westinghouse methods, show that the DNPS and QCNPS can operate at all statepoints, including the minimum core flow statepoint, without violating their SLMCPR in the event of an anticipated operational occurrence . The minimum core flow statepoint SLIVICPR calculations should demonstrate that DNPS and QCNPS can operate at the minimum flow statepoint with some margin Response to Part_

The generic SLMCPR methodology is described in Section 5.3 .2.1 and the Response to RAI's F11 and F13 of Reference 5. The methodology was further clarified in the Response to RAI D-13 of Reference 6.

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The requirement to calculate the SLMCPR at the worst case conditions for minimum core flow conditions at rated power is covered by the general requirement that "the SLMCPR is established based on a single conservative radial power distribution used to represent that cycle" in Section 5.3.

]',' While CPR increases at reduced core power r

relative to rated conditions, A is necessary to specifically evaluate single-loop conditions since I An example of application of the methodology is provided in the response to Part c which is an outline of the QCNPS Unit 2 Cycle 19 SLMCPR analysis.

Response to Part b Since the SLMCPR is based on the number of fuel rods expected to be in boiling transition, the SLMCPR increases as the number of assemblies with CPRs close to the limiting CPR assembly and the number of fuel rods with CPRs close to the limiting fuel rod CPR increase .

Consequently, the SLMCPR increases as the relative assembly and fuel rod power (and, therefore, CPR) distributions become more uniform. Only an assembly at the OLMCPR has the potential to challenge the SLMCPR during an AOO. [

Response to Part c The SLMCPR analysis for QC2, Cycle 19, is currently in progress, and the scope of that analysis provides an illustration of the process described above .

Two-hoop Since the SLMCPR will be the interplay of various factors (e.g . assembly power and fuel rod power distributions), it is calculated throughout the Reference Core cycle to deter conservative SLMCPR as follows.

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Based on the results of Step a, additional state points are evaluated to find the most limiting point(s) in the cycle .

C. At the most limiting cycle burnup(s),

PI C Single Loop I

P,c Additional points any be evaluated as required to establish the t credible le-loop SLMCPR for the cycle .

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INIRC Request 20

.4 of Attachment 7 does not provide sufficient information regarding the Stability Analysis for the staff to reach a safety determination. The staff expects the following documentation to be submitted in a supplemental submittal to the TS Amendment that was previously reviewed by the staff:

a. Provide a summary of the process followed by Westinghouse and plants with house fuel to implement-Long Term Stability Solution Ill.

b. Provide a summary of the process followed by Westinghouse to calculate plant-specific setpoints and core operating limits report items .

c. Provide a list and short description of the major codes used by Westinghouse and their uses for licensing applications .

the status of the licensing basis for these methodologies and identify any topical reports that are NRC-approved or under review to support the methodologies.

e . Document the plant-specific DIVOM calculation for each plant .

Response

a . The process followed by Westinghouse to implement the Long-Term Stability Solution III is the process developed by the BWROG as described in References 10 and 20. In the following table the Westinghouse methodology is compared to the cycle-specific DIVOM procedure guideline:

Table 20-1 Plant Specific DIVOM Procedure Plant Specific Regional Mode DIVOM No . Element Procedure Guideline (Ref. 10) Westinghouse Methodology 1 Plant-specific Generate base deck for plant to be Consistent with the guideline, model evaluated . Westinghouse sets up a RAMONA3 input deck following the procedures established in References I I and 13.

2 'Cycle-specific Incorporate cycle-specific characteristics Consistent with the guideline, model into base deck (bundle types, CPR Westinghouse sets up a steady-state correlation, etc .) . operating conditions following the procedures established in References 11 1 and 13.

3 .3D simulation Generate best-estimate steady-state Consistent with the guideline, data neutronic and thermal-hydraulic data with a Westinghouse generates best-estimate i 3D methodology at the desired powertflow steady-state neutronic and thermal-state point . hydraulic data using the 3D POLCA code (Ref . 14) and procedures established in References 10 and 12 .

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Plant Specific Regional Mode DIVOM I No . Element Procedure Guideline (Ref. 10) Westinghouse methodology 4 Channel ~Optionally, group channels based on Westinghouse models the entire core such grouping established criteria (e .g ., channel power) . that each channel is represented . Different The least stable channel should be stability modes can therefore be evaluated

!considered . The first harmonic flux directly. Therefore, there is no need to

~distribution needs to be computed for artificially group channels.

'regional mode oscillations.

5 ~Cycle 'Generate 3D simulation data for a minimum Consistent with the guideline,

~exposure `of three exposures, e.g ., BOC, PHE, and Westinghouse generates 3D simulation EOC exposures, analyzed at NC at the data for at least three exposure conditions, highest licensed rod line . Nominal rod at the analytical NC conditions along the patterns are used for each exposure. highest licensed rod line .

6 Power/flow Powertflow state points along the highest Consistent with the guideline, conditions Trod line (limited to MELLLA) beginning with Westinghouse simulates conditions along

!NC, then NC+5%, NC+10%, etc., until the highest licensed rod line beginning at

!oscillations tail to develop or the slope of the calculated NC condition and in 5%

1 DIVOM data decreases with increased flow. increasing flow increments until oscillations fail to develop or the slope of the DIVOM data decreases with increased flow.

7  !Xenon Use rated core power equilibrium xenon. Consistent with the guideline, the condition Westinghouse methodology uses rated care ;

power equilibrium xenon 8 ~~Feedwater Use off-rated equilibrium temperature Consistent with the guideline, the i temp (nominal feedwater heating) . Westinghouse methodology uses off-rated I equilibrium feedvvater temperature .

9 !Radial peaking Include consideration for changes in radial Consistent with the guideline, the factor of peaking from the design calculations . The Westinghouse methodology increases the limiting goal is to reasonably represent expected hot bundle power in order to cover expected];

channel. variations in radial peaking factor as the cycle variations .

result of normal operation .

10 Transient Run the transient until the MCPR equals Consistent with the guideline, the simulation 1 .00, until the oscillations are no longer Westinghouseuntil methodology runs the increasing, or until sufficient information is transient the MCPR equals 1 .00, until obtained to generate a DIVOM. the oscillations are not longer increasing, or until sufficient information is obtained to generate a DIVOM.

11 !Ulvulvl Compute (initial- minimum)/initial CPR as a Consistent with the guideline, the calculation function of (peak-minimum)/average Westinghouse methodology computes oscillation magnitude. Connect data to points of ACPR/initial CPR as a function of generate piecewise linear curve. oscillation magnitude for a representative group of hot channels. A point on the DIVOM curve is established by the channel producing the highest ACIDR/initial CPR and :

the channel producing the highest oscillation magnitude. The points are connected to form a piecewise linear curve, inning of Cycle (BOC), Peak Hot Excess (PHE), End of Cycle (EOC) and Natural Circulation (NC) .

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led because best-est The Westinghouse methodol bly lim uideline .

out service, a Backup Stability Protectio In the event the is of results of thi are the locations of Ties on the power-flow s is is perfo, Westinghou incorporate the following s is used in order to verify that the SCRAM REGION and the EXIT R EGI T chosen. The Westinghouse methodology is consistent with the

19) .

The IDN PS and QC N PS stability analysis process is described below for more clarification.

The DNPS and QCNPS reactors are using Long Term Stability Solution Option 111 . The generic approach is currently changed to a cycle specific approach (defined by BVVROG Guideline of Reference 10). The previous fuel vendor has performed the first station Option III e.g. for QCNPS Unit2 (QC2). Starting with

) reload, Wes house will perform the Option III stability analysis to establish the stability based operat limit MCPR as a function of oscillation power range moni mplitude setpoint . Also, Welting rm the Protection (BSP) analysi le 19. Westinghouse has performed a BSP evaluation for a representative first trap reload of SVEA-96 Optima2 fuel in I The result ver ting exclusion zones reasonable margin . Also, stinghouse ?d DIVOM calculation in sup Option III evaluation for C cc I

There are no interactions between Westinghouse and the previous fue on evaluation of stability for 7016-NP.doc Page 42 of 46 NFBEX-0&15 NP-Attachment

b. Westinghouse provides cycle-specific information to support the OPRM setpoint and power-flow map exclusion boundaries to support continued operation should the OPRM be out of service.

The process followed to calculate the cycle-specific DIVOM curve is described in the response to item 'a'above. The OPRM setpoint is established or confirmed to ensure that oscillations initiated following a two-pump trip or steady-state operation at [

]". In the first scenario, the initial MCPR is the MCPR that exists after the coast down to natural circulation and after the feedwater temperature reaches equilibrium. It is assumed that the reactor was operating at the MCPR operating limit prior to the two recirculation pump trip. In the second scenario, the plant is assumed to be in steady-state oiler t

]a,c . It is assumed that the reactor is operating at the MCPR operating limit corresponding to the specified power and flow conditions .

The process followed to determine the power-flow map exclusion boundaries, in the event the OPRM is out of service, is described in the response to item `a' above.

For DNPS and QCNPS, Westinghouse will produce constant decay ratio lines and OPRM trip setpoint versus confirmation count setpoint in support of Option III stability analysis . This analysis is being verified and is scheduled to be completed and available for the N RC review by February 15, 2006 .

c. The major codes used by Westinghouse and their uses for licensing applica ns involving stability are shown in the following table:

Table 20-2 Major Westinghouse BWR Stability Analysis Codes Topical Report and Code Description NRC SER PHOENIX4 This 2D lattice code is used to produce nuclear cross section Refs 15 and 16 dependencies that are used by RAIVIONA3 during transient conditions.

POLCA7 This is he steady-state 3D simulator code used to produce 3D Refs 15 and 16 burnup and Xenon distributions that are used by RAIVIONA3.

,RAMONA3 This code is a time-domain 3D transient code . Refs. 11, 12, 13 and 14 For backup stability protection calculations, the code is used to determine the limiting exposure point during the cycle with regard to decay ratio. At the limiting exposure point, the code is then used to determine exclusion boundaries on the power-flow map .

For DIVOM calculation, this code is used to generate regional power oscillations, determine the transient oscillation magnitude of these oscillations and to provide boundary conditions for BISON hot channel calculation .

BISON IFor the DIVOM calculation, this code is used to calculate the Refs. 17 and 18 ICPR variations during the regional power-flow oscillations in Iselected assemblies .

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The configuration control of the codes is made according to Westinghouse standard where code release notes are the main tool.

The Westinghouse stability methods have originally been licensed as described by CENPD-294-P-A and CENPD-295-FvA as well as CENPD-300-P-A Topical Reports. The CPR application (BISON) has been licensed as described in CENPD-292-P-A. The new application codes (PHOENIX4 and POLCA) initiated a re-evaluation of the stability validation . This re-evaluation contains the previously used jet pump specific stability measurements [ ]',' as well as new measurements in 1 1" (cycles 13 and 19) . Here, QC2 Cycle 19 is a first reload of SVEA-96 Optimal fuel for a plant licensed under EPU/MELLLA . This new validation confirms the ability of the RAMONA code to predict stability for part-length fuel at increased operating domains .

d . The licensing bases for the methodologies used by Westinghouse are presented in the table shown in the response to question c above . As shown in the table, all of the methodologies that are used in the stability calculations have been reviewed and approved by NRC .

e . QCNPS Unit 2 (QC2) has armed the plant Oscillation Power Range Monitor (OPRM) and is currently using long term stability solution Option 3 as the primary protection against damaging oscillations . The plant-specific DIVOM calculation for the Quad Cities and Dresden units will be done as part of the cycle-specific reload analysis . EGC and Westinghouse are performing the regional mode DIVOM analysis required to confirm that the OPRM setpoints, provide protection of the plant MCPR safety limit for anticipated oscillations using the approved methodology established in NEDO-32465-A, Reference 20. The cycle-specific confirmation is being performed in accordance with the procedure guideline documented in OG04-01530-260, Reference 10, which was developed jointly by the BWROG Detect and Suppress Methodology Committee, GNF, AREVA, and Westinghouse. For each reload Westinghouse will determine the OPRM trip setpoint versus maximum confirmation count setpoint in support of Option III stability analysis.

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References 1 Letter from P. R . Simpson (Exelon Generation Company, LLC) to U . S . NRC, "Request for License Amendment Regarding Transition to Westinghouse Fuel," dated June 15, 2005 .

WCAP-1 5942-P, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors Supplement 1 to CENP-287 : October 2004 .

LTR-NRC-05-35, "Transmittal Letter to NRC of Responses to NRC Request for Additional Information on WCAP-15942-P Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors Supplement 1 to CENP-287 ."

4. CENPD-288-P-A, "ABB Seismic/LOCA Evaluation Methodology for Boiling Water Fuel,"

July 1996 .

CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel,"

July, 1996.

6. WCAP-1 6081 -P-A, "10x10 SVEA Fuel Critical Power Experiments and CPR Correlations :

SVEA-96 OPTIMA2," March, 2005 .

7. N EDO-32961, "Revision 1, Safety Analysis Report for Quad Cities 1 & 2 Extended Power Uprate," August 2001 .

N U REG-0800, U .S. N RC Standard Review Plan, Section 4.2 Appendix A, June, 1987.

9. ASME Boiler and Pressure vessel Code, Section 11, Part D, Appendix 2, 1992 Edition.

10. OG04-01530-260, "Plant-Specific Regional Mode DIVOM Procedure Guideline," June 15, 2004 .

11 . CENPD-294-P-A, "Thermal-Hydraulic Stability Methods for Boiling Water Reactors,"

July 1996.

12 . "Acceptance for Referencing of ABB/CE Topical Report CENPD-294-P : Thermal Hydraulic Stability Methods for Boiling Water Reactors (TAC No. M92883)," February 22, 1996.

13. CENPD-295-P-A, "Thermal-Hydraulic Stability Methodology for Boiling Water Reactors,"

July 1996 .

14 . "Acceptance for Referencing of ABB/CE Topical Report CENPD-295-P: Thermal Hydraulic Stability Methodology for Boiling Water Reactors (TAC No. M93648)," February 22, 1996.

15 . CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors," December 2000 .

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"Acceptance for Referencing of CENPD-390-P, The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors' (TAC No. MA5659)," July 24, 2000,

17. ENPD-292-P-A, "BISON -One Dimensional Dynamic Analysis Code for Boiling Water Reactors : supplement 1 to Code Description and Qualification," July 1996,

18. CENPD-292-P: "BISON -One Dimensional Dynamic Analysis Code for Boiling Water Reactors: Supplement I to Code Description and Qualification," (TAC No . M90165),"

October 16, 1995 .

19, OG 02-0119-260, "BWR Owner's Group Guidelines for Stability Interim Corrective Action ."

20 . NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996 .

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