ML18130A367

APR1400 PLUS7, Fuel Design for APR1400, Topical Report Safety Evaluation Redacted

ML18130A367
Person / Time
Site:	PROJ0009
Issue date:	05/25/2018
From:	George Wunder NRC/NRO/DNRL/LB2
To:
Wunder G / 415-1494
Shared Package
ML18130A366	List: ML18130A367 ML18130A370
References
APR1400-F-M-TR-13001 Rev. 1
Download: ML18130A367 (39)

Text

OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NEW REACTORS TOPICAL REPORT (TR) APR1400-F-M-TR-13001, REVISION

1. "PLUS7 FUEL DESIGN FOR THE APR1400

," KOREA HYDRO & NUCLEAR POWER CO. LTD PROJECT NO. PROJ0782

1.0 INTRODUCTION

AND BACKGROUND By letter dated September 17, 2013 (Reference 1), as supplemented by letters dated June 26, 2014 (Reference 2), July 21, 2015 (Reference 3), and July 31, 2017 (Reference 4

), Korea Hydro & Nuclear Power Co. LTD (the applicant) requested review and approval of Topical Report (TR) APR140 0-F-M-TR-13001 , "PLUS7 Fuel Design for the APR1400

." By letter dated August 11, 2017 (Reference 5)

, the applicant submitted a revision to the TR which include d updates to address the U.S. Nuclear Regulatory Commission (NRC) staff's requests for additional information (RAIs)

. This TR describes the PLUS7 assembly mechanical design for the applicant's APR1400 nuclear steam supply system (NSSS).

2.0 REGULATORY EVALUATION

Regulatory guidance for the review of fuel system designs and adherence to applicable General Design Criteria (GDC) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix A, "General Design Criteria for Nuclear Power Plants,"

is provided in NUREG

-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP), Section 4.2, "Fuel System Design" (Reference 6). These regulations include:

1. 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light

-water Nuclear Power Reactors," 10 CFR 50.34, "Contents of applications; technical information," and 10 CFR 50.67, "Accident source term" as they relate to the cooling performance analysis of the emergency core cooling system (ECCS) using an acceptable evaluation model and establishing acceptance criteria for light

-water nuclear power reactor ECCSs.

2. General Design Criterion (GDC) 10, "Reactor Design,"

as it relates to assuring that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of AOOs.

3. GDC 27, "Combined Reactivity Control Systems Capability,"

as it relates to control rod insertability under postulated accident conditions.

4. GDC 35, "Emergency Core Cooling," as it relates to the reactor fuel system being designed such that the performance of the Emergency Core Cooling System will not be compromised following a postulated accident.

5. 10 CFR 52.47(b)(1), which requires that a DC application contain the proposed inspections, tests, analyses, and acceptance criteria (ITAAC) that are necessary and sufficient to provide reasonable assurance that, if the ITAACs are performed and the acceptance criteria are met, a plant that incorporates the design certification is built and OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION will operate in accordance with the design certification, the provisions of the Atomic Energy Act, and the NRC's regulations.

In accordance with SRP Section 4.2, the objectives of the fuel system safety review are to provide assurance that:

1. The fuel system i s not damaged as a result of normal operation and anticipated operational occurrences (AOOs), 2. Fuel system damage is never so severe as to prevent control rod insertion when it is required, 3. The number of fuel rod failures is not underestimated for postulated accidents, and

4. Coolability is always maintained.

An approved mechanical design methodology is utilized to demonstrate compliance with the NRC regulations identified in SRP 4.2 Section II.

The staff's objectives for the review of TR APR1400-F-M-TR-13001 are to ensure that the fuel mechanical design methodology adequately addresses SRP Section 4.2 criteria. In addition, based upon Lead Test Assemblies (LTAs) and Commercial Surveillance Assemblies (CSAs), post

-irradiation examinations (PIEs), mechanical testing, past operating experience of similar designs and materials, and fuel performance code predictions, the staff reviewed expected performance of the PLUS7 assembly to ensure it satisfied these objectives. The staff's review is similar in scope to past reviews, e.g. CE (Combustion Engineering) 16x16 NGF fuel assembly design (Reference 7). 3.0

SUMMARY

OF TECHNICAL INFORMATIO N The applicant jointly developed the PLUS7 fuel assembly design with Westinghouse to improve upon the Guardi an fuel performance. This TR describes the design evaluation of the PLUS7 fuel assembly based on in

-reactor performance data and calculations based on previously approved fuel analysis codes and methods.

The general approach of the TR is based on the guidance provided in SRP Section 4.2. In Sections 2 , "Fuel Assembly and Components Design," and 3 , "Fuel Rod Design," of the TR , the applicant present ed design bases, criteria, and evaluations for potential failure mechanisms. The staff summarized these in Section 4.0 , "Technical Evaluation," of this safety evaluation report (SER) as part of the staff's review. While the applicant analyzed and presented the majority of the failure mechanisms identified in SRP Section 4.2 in this TR, there are a few that are not. The applicant addressed failure mechanisms such as fuel seismic response and core coolability related failure mechanisms (e.g. fuel rod ballooning, cladding embrittlement, bursting, etc.) in other topical and technical reports. For these items, the applicant present ed the basis and criteria for these failure mechanisms in this TR but refers the reader to other reports for the evaluation.

Section 3.3 of the TR discusses the codes and methods used to perform the analyses. The applicant presented th e applicability of each code to the PLUS7 and APR1400 designs as well as the compliance with the applicable SERs. Revision 1 to this TR include s discussions and analyses regarding the impacts of burnup dependent Thermal Conductivity Degradation (TCD) in Sections 3.4 and 3.5.

This information supplements some of the information presented in other sections. For example, Section 3.2.2 OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION discusses the design basis, criteria, and evaluation for cladding strain, but Section 3.4.2 later presents additional information regarding the impact of TCD on strain. Therefore, in order for the reader to understand the applicant's analysis of strain for the PLUS7 fuel assembly as a whole, both Section 3.2.2 and Section 3.4.2 must be reviewed.

The applicant present ed a summary of the PLUS7 fuel operational experience in Section 4 , "PLUS7 Fuel Experience

," to support the analyses in Sections 2 and 3.

4.0 TECHNICAL EVALUATION

The staff's review of TR APR1400

-F-M-TR-13001 , Revision 1

, is summarized below:

Verify that the fuel assembly component and fuel rod design criteria are consistent with acceptance criteria identified in SRP Section 4.2 or are otherwise justified.

Verify that the fuel mechanical design methodology is capable of accurately or conservatively evaluating each component with respect to established design criteria.

Verify that the operating experience data database (in

-reactor residence, post

-irradiation examinations, and out

-of-pile testing) supports the operating limits being requested and provides reasonable assurance that no anomalous behavior will occur during batch implementation.

The layout of this SE R closely follows that of TR APR1400

-F-M-TR-13001. 4.1 Fuel Assembly and Components Design Section 2 of TR APR1400-F-M-TR-13001 provides a description of the PLUS7 fuel assembly with comparison s to Guardian and RFA fuel.

Throughout this section

, the applicant made comparisons between calculated physical phenomena and data obtained from LTAs and CSAs. 4.1.1 Fuel Assembly 4.1.1.1 Fuel Assembly Structural Integrity The staff reviewed the fuel assembly structural integrity design basis, criteria, and evaluation as presented in Section 2.2.2.1 of APR1400

-F-M-TR-13001 , Revision 1. The staff's review is summarized below

. Basis In Section 2.2.2.1 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]he objectives of the fuel assembly safety are to provide assurance that (a) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs), (b) fuel system damage is never so severe as to prevent control rod insertion when it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always maintained

." This design basis is consistent with SRP Section 4.2. Therefore, the staff finds this basis to be acceptable

.

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- PROPRIETARY INFORMATION Criteria In Section 2.2.2.1 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that the criteria used to evaluate the PLUS7 fuel assembly structural integrity are those taken from Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code; specifically, For normal conditions:

P m S m P m+P b m For accident conditions

P m S m' P m+P b m' Where: P m = primary membrane stress P b = primary bending stress S m = allowable design stress S m' = allowable design strength for the accident conditions Additionally, the methodology described by the applicant use d normal operation and AOO stress limit criteria even under postulated accident conditions in order to preclude plastic deformation for components important to control element assembly (CEA) insertion.

These criteria are consistent with SRP Section 4.2

, Appendix A. Therefore, the staff finds these criteria to be acceptable

. Evaluation The structural integrity of the assembly components are verified in the following sections describing the individual assembly components. The fuel assembly evaluation for seismic and loss of coolant accident (LOCA) loads is performed in accordance with the NRC licensed CE methodology, CENPD-178-P , Revision 1 (Reference 12). The staff issue d RAI 7954, Question 22 (ML15169A118), requesting the applicant to provide clarification regarding the evaluation for seismic and LOCA loads since the TR stated that the analysis was performed in Section 4.2 of the APR1400 design control document (DCD), but Section 4.2 of the DCD stated that the analysis was performed in the TR. In its response to RAI 7954, Question 22 (ML15202A676), the applicant stated that the evaluation is presented in Technical Report APR1400-Z-M-NR-14010-P , "Structural Analysis of Fuel Assemblies for Seismic and LOCA Loading." The staff confirmed that this information was presented in the referenced technical report and that the staff's SER for DCD Section 4.2 would address the staff's evaluation of the fuel assembly structural analysis of external loads. Therefore, the staff's evaluation will not be presented in this SER.

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- PROPRIETARY INFORMATION 4.1.1.2 Rod to Top Nozzle Axial Cle arance The staff has reviewed the fuel assembly structural integrity design basis, criteria, and evaluation as presented in Section 2.2.2.2 of the APR1400-F-M-TR-13001 , Revision 1. The staff's review is summarized below

. Basis In Section 2.2.2.2 of the TR, the applicant state d that , "[i]f a fuel rod were to be fully constrained axially by contact with top and bottom nozzles, large axial load could be generated. This load could result in fuel rod bowing, overstressing of guide thimbles, or overstressing of guide thimble joints to nozzles." This is consistent with SRP Section 4.2 (1)(A)(v). Therefore, the staff finds this basis to be acceptable.

Criteria Section 2.2.2.2 of the TR states that "[t]he axial clearance between the fuel rod and top nozzle shall be maintained during the fuel life time.

" This criterion will preclude interference between the fuel rods and the nozzles and is acceptable.

Evaluation The applicant performed a simple calculation of axial gap between the fuel rod and top nozzle that includes the fuel rod and assembly irradiation growth to ensure that axial clearance is maintained during the fuel lifetime. Additionally, pool side examination data is provided to confirm that there is sufficient gap after three cycles of irradiation.

The staff reviewed the analysis and finds that this evaluation, supported by pool side examination data

, demonstrates compliance with the stated criteria in that there will be rod to top nozzle axial clearance throughout the requested fuel assembly lifetime. 4.1.1.3 Hydraulic Stability The staff has reviewed the fuel assembly structural integrity design basis, criteria, and evaluation as presented in Section 2.2.2.3 of the APR1400-F-M-TR-13001 , Revision 1. The staff's review is summarized below

. Basis In Section 2.2.2.3 of APR1400-F-M-TR-13001 , Revision 1, the applicant stated that , "[s]ince the fuel assembly lift

-off may cause the fuel assembly and in

-core structure failure, the fuel assembly shall not be lifted off during the normal operation. The fuel assembly and fuel rod vibration causing the fuel failure shall not occur over the full range of flow rates of the plant

." This is consistent with SRP Section 4.2(1)(A)(iii) and (1)(A)(vii). Therefore, the staff finds this basis to be acceptable.

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- PROPRIETARY INFORMATION Criteria The applicant provided the following criteria regarding hydraulic stability:

The fuel assembly shall not be lifted off during normal operation.

Fuel rod vibration causing the fuel failure shall not occur over the full range of flow rates.

The fuel rod vibration caused by the cross

-flow between the fuel assemblies shall not result in fretting wear

-induced cladding failure.

These criteria are consistent with the guidance provided in SRP Section 4.2 (1)(A)(iii) and (1)(A)(vii). Therefore, the staff finds these criteria to be acceptable.

Evaluation The applicant state d that the assembly lift off and fuel rod vibration failure criteria are met as demonstrated through prototype testing and operating experience. The staff reviewed the information provided and confirmed that it covered the full range of flow rates and demonstrated that a fuel assembly will not lift off due to the hydraulic loads nor generate abnormal fluid

-induced vibration which could lead to fuel rod failure. The applicant also provided documentation regarding 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> of endurance tests performed in the VIPER high temperature loop on PLUS7 and Guardian fuel to confirm that cross flow

-induced vibration did not produce grid to rod fretting failure. Additional pool side examinations confirmed that there h a s been no fuel rod fretting failure of PLUS7 fuel since its introduction in 2002. The staff reviewed the provided testing and pool side examination information and concludes that the applicant has adequately demonstrated that PLUS7 fuel cladding will not experience fretting wear

-induced failures due to fuel rod vibration caused by cross

-flow. Based on the above staff evaluations, the staff finds that the applicant adequately demonstrated that PLUS7 fuel meets the hydraulic stability criteria and is

, therefore , acceptable.

4.1.1.4 Shipping and Handling Loads The staff has reviewed the fuel assembly structural integrity design basis, criteria, and evaluation as presented in Section 2.2.2.4 of APR1400-F-M-TR-13001 , Revision 1. The staff's review is summarized below:

Basis In Section 2.2.2.4 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he fuel assembly shall withstand loads encountered during normal shipping and handling conditions

." There are no rules or regulations regarding shipping loads as covered by the guidance provided in SRP Section 4.2. However, this basis is consistent with SRP Section 4.2 regarding fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.2.2.4 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criteria as

(1) the fuel assembly shall meet the stress criteria of normal operation with an axial OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION load of 4g acceleration and a lateral load of 6g acceleration that may occur during shipping and (2) the fuel assembly shall meet the stress criteria of normal operation with the maximum load of 4g acceleration that may occur during handling.

SRP Section 4.2 does not specify that stress criteria must be met at a specific acceleration, only that the stress criteria must be met. Therefore, the staff does not make a finding regarding the 4g and 6g acceleration criteria, but does find the underlying stress criteria to be consistent with SRP Section 4.2 and is

, therefore , acceptable.

Evaluation The applicant provided the results of an evaluation which credit s the grid springs to provide initial pre

-loads greater than 4g axially and 6g laterally. Additionally, the applicant performed prototype testing to confirm that the grids would not deform under 6g lateral loading. The staff's evaluation of the grid testing is captured in the SER for APR1400 DCD , Section 4.2.

The staff reviewed the prototype testing presented in the TR and finds that the results and analysis demonstrate that the PLUS7 fuel assembly will not be damaged by 4g axial and 6g lateral loads.

This analysis and evaluation does not cover the cask nor its ability to limit the loads on the fuel assembly, but instead covers the ability of the PLUS7 fuel assembly to withstand 4g axial and 6g lateral loads.

The staff finds that the applicant adequately demonstrated that PLUS7 fuel assemblies can withstand these loads and therefore meets the stress criteria. 4.1.1.5 Mechanical Compatibility The staff has reviewed the mechanical compatibility design basis, criteria, and evaluation as presented in Section 2.2.2.5 of APR1400-F-M-TR-13001 , Revision 1. The applicant's basis for this criterion is that the PLUS7 fuel assembly can be loaded with no interference with reactor internals and maintain mechanical compatibility with core components.

The staff notes that this criterion does not directly correlate to any NRC rules or regulations; however, there would be secondary relationships with other criteria for which the NRC does have clear requirements (e.g. control rod insert ability, clad stress, etc.). Therefore , the staff makes no finding specifically regarding this criterion other than to state that the implementation of this criterion by the applicant supports, but does not individually address

, the other related criteria. The criteria that do correlate to NRC rules or regulations are evaluated in the appropriate sections of this SER. 4.1.2 Fuel Assembly Components 4.1.2.1 Bottom Nozzle The staff finds that the bottom nozzle design has been adequately described in the TR such that an evaluation of the proposed methods may be performed. The following sections detail the staff's review of the applicant's basis, criteria, and evaluation used to analyze the bottom nozzle in terms of

(1) providing structural support of the fuel assembly, (2) preventing fuel rod ejection, (3) maintaining compatibility with the lower support structure, and (4) maintaining compatibility with the instrument.

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- PROPRIETARY INFORMATION 4.1.2.1.1 Structural Support of the Fuel Assembly Basis In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]he bottom nozzle integrity is maintained for lateral and vertical loads from normal, AOOs, postulated accidents, and shipping

& handling conditions

." There are no regulatory requirements regarding shipping loads as covered by the guidance provided in SRP Section 4.2. However, this basis is consistent with the SRP Section 4.2 (1)(A)(i) basis regarding fuel failure.

Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he bottom nozzle shall withstand lateral and vertical loads without exceeding the stress limits described for the assembly structural integrity discussed in Section 2.2.2.1." The staff finds that by complying with the stated criterion, the PLUS7 fuel assembly bottom nozzle would maintain its structural integrity.

Evaluation The applicant performed a finite element analysis to verify the mechanical integrity of the bottom nozzle for the loads from fuel assembly shipping and handling, normal operation, AOOs, and postulated accidents. This analysis showed that for a conservative estimate of loading there is significant margin to the allowable stress limit.

The staff reviewed the analysis presented in APR1400

-F-M-T R-13001 , Revision 1

, and found that it demonstrates that the PLUS7 bottom nozzle complies with the stated criteria and is

, therefore , acceptable.

4.1.2.1.2 Prevention of Fuel Rod Ejection Basis In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he fuel rod shall not be ejected out of fuel assembly

[sic]." Although the SRP does not provide direct guidance regarding fuel rod ejection, the basis presented by the applicant follow ed the overall guidance provided in SRP Section 4.2 (1)(C) regarding coolability and is , therefore , acceptable.

Criteria In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "[t]he bottom nozzle will be designed so fuel rods are not ejected through the nozzle." The staff notes that SRP Section 4.2 does not provide specific guidance regarding the ejection of a fuel rod (as opposed to the guidance provided regarding Rod Control Cluster Assembly (RCCA) ejection), but the proposed criterion does relate to SRP Section 4.2 (1)(C) regarding coolability. T herefore, the staff finds this criterion to be acceptable.

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- PROPRIETARY INFORMATION Evaluation The PLUS7 bottom nozzle is designed to provide an offset in the centerline of the fuel rod array and the nozzle's primary flow hole array to prevent fuel rod ejection.

The staff reviewed the drawings and confirmed that there is no direct path for a fuel pin to fal l through. Therefore, the staff finds that this design feature is acceptable to prevent fuel rod ejection.

4.1.2.1.3 Compatibility with the Lower Support Structure Basis In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he bottom nozzle can mate with the lower support structure without interference

." This basis is consistent with the SRP Section 4.2 basis regarding fuel failure. Therefore

, the staff finds this basis to be acceptable.

Criteria In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he bottom nozzle shall be acceptable with the lower internal by mating between the bottom nozzle and the lower support structure

." The staff notes that SRP Section 4.2 does not present specific guidance regarding bottom nozzle mating with the lower support structure, but the proposed criterion is consistent with SRP Section 4.2(1)(A)(v) regarding prevention of dimensional changes which could cause fuel failures. Therefore, the staff finds this criterion to be acceptable.

Evaluation The applicant evaluated the geometric compatibility using wor st case dimensional tolerances and it was demonstrated that there will be no interference.

The staff has reviewed the applicant's evaluation and finds that it demonstrates the PLUS7 fuel assembly complies with this criterion. 4.1.2.1.4 Compatibility with the Instrument Basis In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[p]roper alignment with the in

-core instrument will assure reliable function of core monitoring and protection.

" This basis is consistent with the SRP Section 4.2 basis regarding fuel failure. Therefore

, the staff finds this basis to be acceptable.

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- PROPRIETARY INFORMATION Criteria In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he bottom nozzle shall be dimensionally compatible with the in

-core instrument while providing adequate cooling

." The staff notes that SRP Section 4.2 does not present specific guidance regarding bottom nozzle maintaining alignment with the in

-core instrument, but the proposed criterion is consistent with SRP Section 4.2(1)(A)(v) regarding prevention of dimensional changes which could cause fuel failures. Therefore, the staff finds this criterion to be acceptable. Evaluation In Section 2.3.1.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant describe d the design of the bottom nozzle that directs the in

-core instrument through the bottom nozzle plenum region. The staff reviewed the design description and the diagram, and the staff finds that the design shows that the bottom nozzle is dimensionally compatible with the in

-core instrument while providing adequate cooling and complies with this criterion. T herefore, the staff finds this criterion to be acceptable.

4.1.2.2 Top Nozzle The applicant provided a description of the top nozzle and associated figures in Section 2.3.2 of APR1400-F-M-TR-13001 , Revision 1. The staff finds that the top nozzle design has been adequately defined in the TR such that an analysis of the proposed methods may be performed.

The applicant identified five areas where analysis should be performed to ensure that the top nozzle meets its design requirements.

The staff's review of these areas is presented below. 4.1.2.2.1 Structural Support of the Fuel Assembly Basis In Section 2.3.

2.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he top nozzle integrity is maintained for lateral and vertical loads from normal, AOOs, postulated accidents, and shipping

& handling conditions

." There are no regulatory requirements regarding shipping loads as covered by the guidance provided in SRP Section 4.2. However, this basis is consistent with the SRP Section 4.2 basis regarding fuel failure. Therefore

, the staff finds this basis to be acceptable. Criteria In Section 2.3.

2.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "[t]he top nozzle shall withstand lateral and vertical loads without exceeding the stress limits of Section 2.2.

2.1". The staff notes that SRP Section 4.2 does not present specific guidance regarding top nozzle structural integrity, but the proposed criterion is consistent with SRP Section 4.2(1)(A)(i) regarding stress, strain, and load limits for fuel assembly components in order to prevent fuel failures. Therefore, the staff finds this criterion to be acceptable.

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- PROPRIETARY INFORMATION Evaluation The applicant performed a finite element analysis to verify the mechanical integrity of the bottom nozzle for the loads from fuel assembly shipping and handling, normal operation, AOOs, and postulated accidents. This analysis showed that for a conservative estimate of loading there is greater than 60 percent margin to the allowable stress limit.

There are no regulatory requirements regarding shipping loads as covered by the guidance provided in SRP Section 4.2. However, the staff reviewed the evaluation presented in the TR and finds that the basis, criterion, and analysis provided demonstrate that the bottom nozzle design compli es with stress limits regarding the fuel failure criterion for normal operation, AOOs , and postulated accidents. 4.1.2.2.2 Prevention of Fuel Rod Ejection Basis In Section 2.3.2.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he fuel rod shall not be ejected out of the fuel assembly

." Although the SRP does not provide direct guidance regarding fuel rod ejection, the basis presented by the applicant follow ed the overall guidance provided in SRP Section 4.2 (1)(C) regarding coolability and is

, therefore , acceptable.

Criteria In Section 2.3.2.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he top nozzle shall be designed fuel rods not to be ejected through the nozzle

[sic]." The staff notes that SRP Section 4.2 does not provide specific guidance regarding the ejection of a fuel rod (as opposed to the guidance provided by RCCA ejection), but the proposed criterion does relate to SRP Section 4.2 (1)(C) regarding coolability. T herefore, the staff finds this criterion to be acceptable.

Evaluation The applicant state d that the top nozzle is designed to provide ligaments that are oriented above the fuel rod centerlines to prevent fuel rod ejection and flow holes that are offset laterally from the fuel rod horizontal position.

The staff reviewed the drawings and confirmed that there is no direct path for a fuel pin to fall through. Therefore, the staff finds that this design feature is acceptable to prevent fuel rod ejection.

4.1.2.2.3 Compatibility with the Upper Guide Structure Basis In Section 2.3.

2.2 of APR1400-F-M-TR-1 3001 , Revision 1, the applicant state d that , "[t]he top nozzle can mate with the upper guide structure without interference

."

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- PROPRIETARY INFORMATION This basis is consistent with the SRP Section 4.2 basis regarding fuel failure. Therefore

, the staff finds this basis to be acceptable.

Criteria In Section 2.3.

2.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he top nozzle shall be acceptable with the upper internal by mating between the top nozzle and the upper guide structure

." The staff notes that SRP Section 4.2 does not present specific guidance regarding top nozzle mating with the upper guide structure, but the proposed criterion is consistent with SRP Section 4.2(1)(A)(v)

, regarding prevention of dimensional changes which could cause fuel failures. Therefore, the staff finds this criterion to be acceptable. Evaluation The applicant evaluated the geometric compatibility using wor st case dimensional tolerances and concluded that there is no interference.

The staff has reviewed the applicant's evaluation and finds that it demonstrates the PLUS7 fuel assembly complies with this criterion because the evaluation confirms that the top nozzle and guide structure would mate without causing interference.

Therefore, the staff finds this to be acceptable.

4.1.2.2.4 Remote Reconstitutability In Section 2.3.2.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant provided a criteri on regarding remote reconstitutability of the fuel assembly. There is no guidance in SRP Section 4.2 regarding remote reconstitutability and it does not directly tie to any regulations listed in SRP Section 4.2. Therefore, the staff makes no findings regarding this criterion and did not perform an evaluation other than to note that it does not directly conflict with any NRC regulations.

4.1.2.2.5 Compatibility with Handling Equipment There is no guidance in SRP Section 4.2 regarding compatibility with handling equipment and it does not directly tie to any regulations listed in SRP Section 4.2. Therefore, the staff makes no findings regarding this criterion and did not perform an evaluation other than to note that it doe s not directly conflict with any NRC regulations.

4.1.2.3 Holddown Spring In Section 2.3.3 and Figure 2

-13 of APR1400-F-M-TR-13001, Revision 1 , the applicant present ed a description of the holddown spring. The staff finds that the holddown spring design has been adequately defined in the TR such that an analysis of the proposed methods may be performed.

The applicant identified three areas where analysis should be performed to ensure that the holddown spring meets its design requirements. The staff's review of these areas is presented below.

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- PROPRIETARY INFORMATION 4.1.2.3.1 Holddown Force in Normal Operation Basis In Section 2.3.

3.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[f]uel assembly lift

-off during normal operating conditions can result in fuel assembly or core support structural damage. If lift-off occurs during normal operation, large amplitude fuel assembly oscillations could ensue. These oscillations can cause impact forces on the core structures and fuel assembly.

" This basis is consistent with the SRP Section 4.2 basis regarding fuel failure. Therefore

, the staff finds this basis to be acceptable.

Criteria In Section 2.3.

3.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he fuel assembly shall not lift

-off the core support structure during the normal operation." The staff notes that this criterion is the same criterion used in Section 2.2.2.3 of APR1400-F-M-TR-13001 , Revision 1

. The criterion is also consistent with the guidance in SRP Section 4.2 (1)(A)(vii). Therefore, the staff finds this criterion to be acceptable.

Evaluation The applicant presented results from prototype testing and operating experience in order to evaluate the PLUS7 fuel assembly against this criterion. The testing results presented in Section A.2.9 of APR1400-F-M-TR-13001, Revision 1 , demonstrate that the lift force for the PLUS7 assembly in the APR1400 reactor design would not result in assembly liftoff. The staff reviewed the information provided and finds that the applicant adequately demonstrated that the PLUS7 fuel assembly will not become unseated during normal operation, and therefore complies with this criterion.

Therefore, the staff finds this to be acceptable.

4.1.2.3.2 Maximum Deflection Range and Solid Condition Basis In Section 2.3.

3.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[l]oad of solid spring may cause improper load distribution through the top nozzle, and may have an adverse effect on the spring itself. The load of solid spring on the fuel assembly can cause extensive structural damage and subsequent fuel failure

." The staff finds that this basis is consistent with the guidance provided in SRP Section 4.2 regarding fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.2.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he clearance between the holddown working height and spring solid height shall be maintained under normal operation

." This criterion follows the guidelines in SRP Section 4.2 (1)(A)(v). Therefore, the staff finds this criterion to be acceptable.

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- PROPRIETARY INFORMATION Evaluation The applicant state d that the holddown working height was calculated based on the maximum fuel assembly growth. This working height was compared to the solid spring height to ensure margin between these values.

The staff reviewed the method used to calculate the holddown working height and solid spring height and finds that the applicant demonstrated that the PLUS7 top nozzle spring does not go solid under normal operation

. Therefore, the staff finds this to be acceptable.

4.1.2.3.3 Holddown Spring Shear Stress Basis In Section 2.3.

3.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]he spring shear stress shall meet the ASME Boiler and Pressure Vessel Code

." The staff finds that this basis is consistent with the guidance provided in SRP Section 4.2 regarding fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.

3.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he spring shear stress for normal operation except lift

-off shall be less than 0.8 S m." This criterion for spring shear stress is consistent with ASME B&PV C ode Section III, NB-3227.2, specifically the second part which specifies that the maximum primary shear stress at the periphery of a solid circular section in torsion shall be limited to 0.8 S

m. Additionally, the criterion includes the words "normal operation except lift

-off." Assembly lift

-off is precluded by the design of the spring as discussed in Section 4.1.1.3 of this SER. The staff finds that this criterion is consistent with the guidance provided in SRP Section 4.2 (1)(A)(i). Therefore, the staff finds this criterion to be acceptable. Evaluation The applicant provided the results of a design calculation that demonstrate margin to the shear stress limit. The staff reviewed the results and finds that the applicant demonstrated that the PLUS7 fuel assembly complies with this criterion.

Therefore, the staff finds this to be acceptable.

4.1.2.4 Guide Thimble and Instrument Tube In Section 2.3.4 , Figure 2-1, and Figure 2

-17 of APR1400-F-M-TR-13001, Revision 1

, the applicant present ed a description of the guide thimbles and instrument tube. The staff finds that the guide thimble and instrument tube spring design has been adequately defined in the TR such that an analysis of the proposed methods may be performed.

The applicant identified two areas where analysis should be performed to ensure that the guide thimbles and instrument tube meet their design requirements. The staff concurs that these two OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION areas are appropriate to analyze the guide thimbles and instrument tubes, and the staff's review of these areas is presented below. 4.1.2.4.1 Structural Support of the Fuel Assembly Basis In Section 2.3.

4.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]he guide thimble maintains its structural integrity for lateral and vertical loads from normal, AOOs, and shipping and handling conditions

." The staff's review, based on the guidance provided in SRP Section 4.2, does not cover shipping and handling loads. However, t he staff finds that the basis is consistent with the guidance in SRP Section 4.2 regarding fuel failure. Therefore, the staff finds this basis to be acceptable for supporting the fuel failure requirements. Criteria In Section 2.3.

4.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that the criterion is the same as the structural design criteria of the fuel assembly described in Section 2.2.2.1 of APR1400-F-M-TR-13001 , Revision 1 for the fuel assembly

. These criteria follow the guidance of Section III of the B&PV Code of ASME. For ease of use, these criteria are repeated below

. For normal conditions:

P m S m P m+P b S m For accident conditions:

P m S m' P m+P b m' Additionally, the methodology described by the applicant use d normal operation and AOO stress limit criteria even under postulated accident conditions in order to preclude plastic deformation for components important to CEA insertion.

These criteri on are consistent with SRP Section 4.2

, Appendix A

. Therefore, the staff finds these criteri on to be acceptable.

Evaluation The applicant provided the results of a design calculation of the guide thimble tube during shipping and handling. The applicant concluded that the shipping and handling loads are greater than the loads during normal operation and AOO. Additionally

, the applicant state d that the circumferential stress acting on the guide thimble tube from the pressure difference between inner and outer guide thimble tube around the dashpot region was evaluated to satisfy the stress design limit and that

, conservatively assuming that the CEA drop occurs once a day, the guide thimble tube stress was calculated to be so small that the fatigue

-induced guide thimble tube failure does not occur.

OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION The staff reviewed the analyses and results presented by the applicant and, for the reasons stated above, finds that the applicant demonstrate d that the PLUS7 fuel assembly complies with the stated criteria. As stated in the Basis section, there are no regulatory requirements regarding shipping loads as covered by the guidance provided in SRP Section 4.2. However, the basis, criterion, and analysis provided do demonstrate compliance with stress limits regarding the fuel failure criterion. Therefore, the staff finds this to be acceptable.

4.1.2.4.2 CEA Drop Time Basis In Section 2.3.4.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[p]lant reactivity control systems must be capable of reducing power within a maximum specified time to preclude fuel damage

." The staff finds that the basis is consistent with the guidance in SRP Section 4.2 regarding fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.

4.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he geometrical shape of the guide thimble tube shall meet the control rod drop time limit

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 (1)(A)(vii) regarding control rod insertability. Therefore, the staff finds this criteri on to be acceptable. The staff notes that this criterion is specific to the design of the guide tubes and not the actual drop time analysis itself which is covered in Chapter 4 of the staff's SER for the APR1400 DCD. Evaluation As stated in the Basis section above, the staff notes that CEA drop times are analyzed as part of the APR1400 DCD , Chapter 4 review. Therefore, the staff makes no specific finding here.

4.1.2.5 Grid In Section 2.3.5.1 of A PR1400-F-M-13001-P , Revision 1, Figures 2

-3, 2-4, 2-5, 2-6, 2-19, 2-20, and 2-21 , the applicant present ed a description of the PLUS7 grids (top grids, bottom grids, protective grids, and mixing vane mid grids). The staff finds that the grid designs ha ve been adequately defined in the TR such that an analysis of the proposed methods may be performed.

The applicant identified six areas where analysis should be performed to ensure that the grids meet their design requirements. The staff finds that these six areas cover the failure mechanisms identified in SRP Section 4.2 regarding fuel system damage. The staff's review of these areas is presented below. 4.1.2.5.1 Fuel Rod Support Basis In Section 2.3.

5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]he grid shall support the fuel rod axially and laterally

."

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- PROPRIETARY INFORMATION The staff finds that the basis is consistent with the guidance in SRP Section 4.2 regarding fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.

5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant define d the applicable criterion as, "

[t]he positive spring force of the bottom grid shall be maintaine d up to EOL." The applicant provide d an additional design criterion regarding the minimum spring force for the PLUS7 bottom grid.

The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 (1)(A)(1) regarding loading limits. Therefore, the staff finds this criterion to be acceptable.

Evaluation In Section 2.3.5.2 of APR1400-F-M-TR-13001, Revision 1 , the applicant describe d an evaluation of the bottom grid spring forces of the PLUS7 fuel and shows that the spring force is larger than the minimum spring force criterion at EOL. The staff has reviewed this information and find s that this evaluation demonstrates that the PLUS7 bottom grid design meets this criterion

. Therefore, the staff finds this to be acceptable.

4.1.2.5.2 Shipping and Handling Basis In Section 2.3.5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]he fuel should be designed to withstand shipping and handling loads without damage

." There are no specific regulatory requirements regarding shipping loads in Title 10 of the Code of Federal Regulations (10 CFR), Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants,"

for the fuel system design, as found in the guidance provided by SRP Section 4.2. However, the inclusion of this criteria does not interfere with regulatory requirements.

Therefore, the staff fin ds this basis to be acceptable, but not necessary in order to support the fuel system design regulatory requirements

. Criteria In Section 2.3.5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]he grid shall withstand 4g axial and 6g lateral accelerations without allowing the fuel rods to shift or the grid to permanently deform during the shipping and handling

." As already noted, there are no regulatory requirements regarding shipping and handling loads within the guidance provided in SRP Section 4.2. However, the use of 4g axial and 6g lateral acceleration limits does not conflict with staff guidance regarding loading limits provided in SRP Section 4.2 (1)(A)(i).

Therefore, the staff finds this criterion to be acceptable.

Evaluation The applicant performed an evaluation to ensure that the friction loads associated with the spring force on the grids is greater than an axial load of 4g and the grid spring force is greater than a lateral load of 6g.

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- PROPRIETARY INFORMATION Additionally, the applicant performed grid buckling tests to demonstrate that the buckling strength of the grids is greater than a lateral load of 6g.

As already noted, the staff has no regulations or guidance regarding shipping loads within 10 CFR Part 52 or SRP Section 4.2 in terms of the fuel system design. Therefore, the staff makes no specific regulatory finding regarding this criterion. 4.1.2.5.3 Fuel Rod Fretting Wear Basis In Section 2.3.5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he fuel rods should not fail due to the grid

-to-rod fretting wear caused by the fuel assembly

-induced or the fuel rod-induced vibration during normal operation

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding loading limits and fuel failure. Therefore , the staff finds this basis to be acceptable.

Criteria In Section 2.3.5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he fuel rod should not fail due to fretting wear during fuel lifetime.

" The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 (1)(A)(iii) regarding fuel rod fretting. Therefore , the staff finds this criterion to be acceptable.

Evaluation The applicant use d prototype testing and operating experience to evaluate the PLUS7 fuel assembly against this criterion

. The applicant state d that fluid

-induced assembly vibration tests have been performed to confirm that the symmetric arrangement of the PLUS7 grid mixing vanes will cause very little fluid

-induced vibration. Furthermore, 500

-hour wear resistance tests show that the PLUS7 fuel will not generate fretting wear

-induced fuel failure for the fuel lifetime. Finally, operating experience demonstrates that the PLUS7 fuel assembly has no t shown any fretting wear failure for the fuel lifetime.

The staff has reviewed the test descriptions and results presented in Section 2.3.5.2 and Appendices A.2.2 and D of APR1400-F-M-TR-13001 Revision 1. The staff finds that Figure 2-10 demonstrates that for the range of flow rates for APR1400, the maximum vibrational amplitude remains small and would not lead to hydraulic instabilities. Additionally, the long

-term fretting wear test results from Appendix A.2.2 demonstrate that no measurable fretting wear occurs on oxidized rods and that the fretting wear on non

-oxidized rods meets KHNP's internal criterion. Appendix D shows through extrapolation and use of a conservative failure criterion, a non-oxidized rod would not experience failure for the PLUS7 fuel assembly in an APR1400 plant design. Based on these tests results and analyses, the staff finds that the applicant demonstrate d that the PLUS7 fuel assembly will not experience fuel rod fretting wear failure over the lifetime of the fuel

. Therefore, the staff finds this to be acceptable.

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- PROPRIETARY INFORMATION 4.1.2.5.4 Fuel Rod Bow Basis In Section 2.3.5.2 of APR14 00-F-M-TR-13001 , Revision 1, the applicant state d that , "-the grids shall accommodate the fuel rod length change caused by the thermal expansion and neutron irradiation growth without inducing unacceptable fuel rod bow

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding rod bowing and fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.5.2 of APR1400-F-M-TR-130 01 , Revision 1, the applicant state d , "[t]he grid shall not permit or cause rod bowing that exceeds the allowable limits for channel closure for the fuel assembly lifetime

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2(1)(A)(v) regarding fuel rod bowing. Therefore, the staff finds this criterion to be acceptable.

Evaluation The applicant perform ed an evaluation to determine if a departure from nucleate boiling ratio (DNBR) penalty is needed for the PLUS7 fuel design based on the expected reduction in fuel channel spacing due to rod bow. The applicant provided results of in

-reactor tests to conclude that the PLUS7 rod bow is less than the limit, therefore no penalty is needed to ensure that the rod bow criterion is met.

The staff reviewed the test results used by the applicant to evaluate rod bow and finds that the applicant demonstrated that the PLUS7 fuel assembly design complies with this criterion.

4.1.2.5.5 Mid Grid Buckling Strength Basis In Section 2.3.5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[c]ore coolability and safe shutdown of the reactor should be maintained under the most limiting load on the mid grid assembly

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2

, Appendix A

, regarding fuel assembly structural response to externally applied forces

. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he grids must provide a coolable geometry and allow the control rod insertion for the postulated accidents, such as seismic and LOCA

."

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- PROPRIETARY INFORMATION The staff has reviewed the information provided and notes that this criteri a is consistent with the guidance provided in SRP Section 4.2

, Appendix A

, regarding determination of strength for grids. Therefore, the staff finds this criteria to be acceptable.

Evaluation The applicant use d a n NRC-approved CE methodology to perform grid strength evaluation. Grid buckling strength was determined from dynamic impact testing for comparison with predicted grid impact force. The details of this analysis are contained in Technical Report APR1400-Z-M-NR-14010-P , Revision 2, and the staff's review is contained in the SER for APR1 400 DCD Section 4.2.

4.1.2.5.6 Grid Width Basis In Section 2.3.5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]he grid width will grow due to neutron irradiation as the fuel burnup increases. This grid irradiation growth should not affect the fuel loading and unloading operation in the reactor core

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding rod bowing and fuel failure. Therefo re, the staff finds this basis to be acceptable.

Criteria In Section 2.3.5.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[a]n adequate space between the fuel assemblies should be maintained to load and unload the fuel assembly safely for the fuel lifetime

." The staff notes that SRP Section 4.2 does not provide specific guidance regarding irradiation induced grid width growth, but this criterion generally supports the guidance provided in SRP Section 4.2 (1)(A)(v) regarding dimensional changes. Therefore, the staff finds this criterion to be acceptable.

Evaluation The applicant's evaluation referred to poolside examination of PLUS7 fuel to confirm that there is adequate space between the fuel assemblies after the three cycle irradiation.

The applicant also credit ed the designs of the grid guide tab, grid guide vane, and grid corner shape of PLUS7 fuel as being based on the previous designs used successfully in commercial nuclear power plants. The staff has reviewed the applicant's evaluatio n and finds that fuel assemblies at EOL conditions will not experience growth to an extent which would allow neighboring fuel assemblies to interfere with each other.

The staff finds that the PLUS7 fuel assembly design meets this criterion

. Therefore, the staff finds this to be acceptable. 4.1.2.6 Joint and Connection Five joint and connections are identified for the PLUS7 fuel assembly. These include; top nozzle/guide thimble, top grid/guide thimble/instrument tube, mid grid/guide thimble/instrument OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION tube, bottom grid/guide thimble/instrument tube, and bottom muzzle/protective grid/guide thimble. These joints and connections have been adequately defined in the TR such that an analysis of the proposed methods may be performed.

In Section 2.3.6.1 of APR1400-F-M-TR-13001 , Revision 1

, and Figures 2

-22, 2-23, and 2-25 , the applicant present ed a description of the PLUS7 fuel assembly joints and connections. The staff finds that the joints and connections have been adequately defined in the TR such that an analysis of the proposed methods and analysis can be performed.

The applicant identified five areas where analysis should be performed to ensure that the joints and connections meet their design requirements. The staff agrees that these five analysis areas are sufficient to demonstrate compliance. The staff's review of these areas is presented below. 4.1.2.6.1 Top Nozzle/Guide Thimble Basis In Section 2.3.6.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he top nozzle/guide thimble joint and connections shall not be damaged under the loads of normal operation and AOOs

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding stress, strain, and loading limits which can lead to fuel failure

. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.

6.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d: Under the loads of normal operation and AOOs, the top nozzle/guide thimble joint and connections shall not be damaged and must meet the following design criteria:

P m y and 2/3 S u y Where, P m = calculated primary membrane stress S u = minimum ultimate tensile strength at unirradiated condition S y = minimum yield tensile strength at unirradiated condition The staff has reviewed the information provided and notes that this criterion is consistent with ASME Section III

, Division 1

, Subsection NG

-3232.1 for threaded structural fasteners. Therefore, the staff finds this criterion to be acceptable.

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- PROPRIETARY INFORMATION Evaluation The applicant provided results from a design calculation which demonstrate that the stress limits will not be exceeded. The staff ha s reviewed the applicant's evaluation and finds that the PLUS7 fuel assembly meets this criterion

. Therefore, the staff finds this to be acceptable.

4.1.2.6.2 Grid/Guide Thimble/Instrument Tube Basis In Section 2.3.6.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[d]imensional stability of the fuel assembly must be maintained under operating, shipping, and handling conditions. The joint must remain intact under operating, shipping, and handling conditions. Since the spot

-welds are not expected to yield significantly before failure, the design limits are based on failure loads

." There are no regulatory requirements regarding shipping loads as covered by the guidance provided in SRP Section 4.2. However, this basis is consistent with the SRP Section 4.2 basis regarding stress, strain, and loading limits which can lead to fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.

6.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant stat ed , "[u]nder the loads of normal operations and AOOs

, the grid/guide thimble/instrument tube joint and connections shall not be damaged

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 (1)(A)(1) regarding loading limit

s. Therefore, the staff finds this criterion to be acceptable. Evaluation The applicant state d that the largest load acting on the grid/guide thimble welding point is a slip load at the fuel rod/grid interface caused by the difference in their thermal expansion. The applicant state d that it calculated this stress and that it was evaluated to be far less than the weld strength of the grid/guide thimble connection measured by test.

The staff reviewed this evaluation and found that it is acceptable in identifying the largest load and in calculating the stress due to this load and demonstrates that the PLUS7 fuel assembly complies with this criterion

. The staff finds this to be acceptable.

4.1.3.2.3 Bottom Nozzle/Guide Thimble Basis In Section 2.3.6.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[f]uel assembly dimensional stability must be maintained under shipping, handling, operating, and accident conditions. The lifting force requirement ensures that the joint can withstand the maximum possible force exerted on the fuel assembly by the fuel handling equipment during removal from the core.

"

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- PROPRIETARY INFORMATION The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding stress, strain, and loading limits which can lead to fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 2.3.6.2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d: Under the loads of normal operation and AOOs, the top nozzle/guide thimble joint and connections shall not be damaged and must meet the following design criteria:

P m y and 2/3 S u y Where, P m = calculated primary membrane stress S u = minimum ultimate tensile strength at unirradiated condition S y = minimum yield tensile strength at unirradiated condition There are no regulatory requirements regarding shipping loads as covered by the guidance provided in SRP Section 4.2. However, this basis is consistent with the SRP Section 4.2 basis regarding fuel failure for normal operation and AOOs. The staff has reviewed the information provided and notes that this criterion is consistent with ASME Section III

, Division 1

, Subsection NG-3232.1 for threaded structural fasteners. Therefore, the staff finds this criterion to be acceptable.

Evaluation The applicant provided results from a design calculation which demonstrate that the stress limits will not be exceeded. The staff reviewed the applicant's evaluation and find s that the PLUS7 fuel assembly meets this criterion.

Therefore, the staff finds this to be acceptable.

4.2 Fuel Rod Design In Section 3 of APR1400-F-M-TR-13001 , Revision 1, the applicant described the 16 identified design bases that have been identified for the thermal

-mechanical performance of PLUS7 fuel. For each of these design bases, the applicant described the criteria and evaluation used to show that the PLUS7 fuel would meet these bases. The following sections describe the staff's review of the acceptability of each of these design bases, criteria, and evaluations for thermal

-mechanical performance of PLUS7 fuel.

4.2.1 Fuel Design Bases, Criteria, and Evaluation The staff has reviewed the information provided by the applicant in the TR and finds that the fuel rod design has been adequately defined such that an analysis of the proposed methods can be performed.

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- PROPRIETARY INFORMATION Typically, an applicant requests that a methodology for showing compliance with the fuel rod design bases on a cycle

-specific basis be reviewed and approved. In Revision 1 of APR1400-F-M-TR-13001 , the applicant d id not present a specific methodology to perform these analyses on a cycle

-specific basis, but rather show ed that within a given operating envelope, the design bases will be met. The operating window that has been requested by the applicant is as follows: Linear Heat rate specified acceptable fuel design limit (SAFDL) of

[

] for normal operation and AOO

. Radial fall

-off shown in Figure 1 for UO 2 and Gd 2 O 3-UO 2 fuel rods where powers are normalized to

[

]. Maximum radial peaking factor of

[

]. Maximum fuel rod average burnup of 60 GW D/MTU (Gigawatt-days per metric ton of uranium). Figure 1: Rod Power History Used for PLUS7 Fuel Rod Performance Analysis (powers normalized to

[

]) The primary code that the applicant use d to calculate fuel performance parameters is FATES3B. This code was developed in the 1970s and 1980s and does not include the effect of burnup on fuel thermal conductivity and therefore will under predict fuel rod temperatures with increasing burnup. The applicant has determined that some fuel rod criteria will not be impacted by the lack of burnup dependent fuel TCD such as cladding corrosion and hydrogen pickup, cladding collapse, and fuel rod growth. In these cases

, FATES3B and the associated methods may be used to demonstrate compliance with the design bases. However, other OFFICIAL USE ONLY

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- PROPRIETARY INFORMATION criteria such as cladding stress, cladding strain, cladding fatigue, rod internal pressure, and overheating of fuel pellets are affected by TCD. In these cases

, an alternative code and/or method will be used to demonstrate compliance within the requested operational window. The applicant present ed the results of the burnup dependent TCD analysis in Section 3.4 of APR1400-F-M-TR-13001 , Revision 1, with results that sometimes supersede analyses without TCD earlier in the TR. For clarity, the staff's safety evaluation will include the analysis of the effects of TCD directly in the evaluations of cladding stress, cladding strain, cladding fatigue, rod internal pressure, and overheating of the fuel pellets, instead of handling TCD separately.

4.2.1.1 Cladding Stress Basis In Section 3.2.1 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[a] fuel system will not be damaged due to excessive stress under normal operation including AOOs

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding stress, strain, and loading limits which can lead to fuel failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 3.2.1 of APR1400-F-M-TR-13001 , Revision 1, the applicant provide d two criteria by which to evaluate cladding stress:

During normal operation and AOOs, primary tensile stress in the clad and the end cap welds must not exceed [ ] at the applicable temperature.

During Normal operation and AOOs, primary compressive stress in the clad and end cap welds must not exceed

[ ] at the applicable temperature.

The applicant further specifie d that , "-the unirradiated cladding yield strength is conservatively used as the cladding stress design criterion." The staff has reviewed the information provided and notes that this criterion is consistent with ASME Section III of the B&PV Code. The staff finds that by following the ASME Section III stress limits, these criteria are consistent with the guidance provided in SRP Section 4.2 (1)(A)(i), and are , therefore , acceptable. Additionally, the staff agrees that the use of unirradiated cladding yield stress is conservative for this analysis. Therefore, the staff finds this criterion to be acceptable.

Evaluation In Section 3.2.1 of APR1400-F-M-TR-13001 , Revision 1, the applicant summarized analyses used to evaluate the PLUS7 fuel design against the clad stress criteria. Design calculations were provided that demonstrate that the primary tensile and compressive stresses in the cladding and end cap welds are within the allowable limits.

Additionally, in Section 3.4.1 of APR1400-F-M-T R-13001 , Revision 1, the applicant discusse d the impact of burnup dependent TCD on cladding stress.

The applicant state d that the cladding OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION stress analysis is unaffected by TCD since rod internal pressure is also not affected by TCD (per Section 3.4.4 of t he TR). The staff reviewed the statements regarding TCD and agrees that cladding stress would only be affected by TCD if rod internal pressure is affected since it is the primary cause of cladding stress. As detailed in Section 4.2.1.5 of this SER, the staff finds that rod internal pressure is not affected by FATES3B's lack of a burnup dependent TCD model, and therefore

, the staff agrees that the analysis methods used by the applicant to evaluate cladding stress remain valid.

The staff reviewed the results of the applicant's cladding stress analysis and finds that the PLUS7 fuel design complies with the clad stress criteria, because the primary tensile and compressive stresses in the cladding and end cap welds are within the allowable limits. The analysis and the staff's finding are based on the operating window as discussed in Section 4.2.1 of this SER

. 4.2.1.2 Cladding Strain Basis In Section 3.2.

2 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that the , "[f]uel system will not be damaged due to excessive strain under normal operation including AOOs

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding cladding strain limits which can lead to fuel rod failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 3.2.

2 of APR1400-F-M-TR-13001 , Revision 1, the applicant provide d three criteria by which to evaluate cladding strain: (1) the permanent (plastic) hoop strain will remain below [ ] relative to beginning of life dimensions, (2) the total change in hoop strain (elastic plus plastic) from a single AOO will remain below

[ ] relative to the pre

-transient dimensions , and (3) the applicant state d that they expect ZIRLO to have greater than [ ] ductility at burnups greater than 60 GWd/MTU. The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 (1)(B)(vi). Therefore, the staff finds this criterion to be acceptable.

Evaluation In Section 3.2.1 of APR1400-F-M-TR-13001 , Revision 1, the applicant performed an evaluation of hoop strain for normal operation and AOOs using the methodology from the previously approved TR CENPD-404-P-A, "Implementation of ZIRLO Cladding Material in CE Nuclear Power Fuel Assembly Design

," (Reference 10). The staff noted and independently confirmed that TCD (which is not included in FATES3B) has a significant impact on the prediction of total hoop strain and strain increment. The staff issued RAI 7954, Question 18 (ML15169A118)

, requesting the applicant to provide a sample calculation showing the total strain increment resulting from a typical AOO at 0, 20, 40, and 60 GWd/MTU. The staff performed confirmatory calculations using FRAPCON to compare with KHNP's response (ML17223B382) and are shown in Figure 2. In its response to RAI 7954, OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION Question 18 (ML17223B382), the applicant showed that the FATES

-3B calculation of transient hoop strain is always greater than those calculated using FRAPCON.

As detailed in Section 3.4.2 of the TR , the applicant re-performed these strain calculations to account for burnup dependen ce using FATES3B with the modified NFI fuel thermal conductivity model from FRAPCON. The staff notes that since FATES3B is not tuned to this model, this should result in an overestimate of cladding strain. The resulting cladding plastic strain was

[

] and the total (elastic+plastic) strain was

[

] which still showed margin to the limit of

[

] strain. Figure 2: Comparison of the applicant and FRAPCON calculated cladding total hoop strain

. Based on the applicant's revised analysis which accounts for burnup dependent TCD, the staff finds that the PLUS7 fuel assembly meets the strain criteria.

4.2.1.3 Cladding Fatigue Basis In Section 3.2.3 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that the , "[f]uel system will not be damaged due to excessive fatigue under normal operation

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding loading limits and fuel failure

. Therefore , the staff finds this basis to be acceptable.

Criteria In Section 3.2.3 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[f]or the number and type of transients which occur during normal operation, end

-of-life (EOL) cumulative fatigue damage in the cladding must be less than

[ ]." Additionally, the applicant provide d further description of the cladding fatigue criteria by stipulating that the analysis is performed with the Langer-O'Donnell fatigue design curve and a safety factor of 2 on the stress or a safety factor of 20 on the number of cycles is imposed on the curve.

OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION The staff notes that this criterion is conservative as fatigue failure is typically predicted at a fatigue damage fraction of 1.0 and the Langer

-O'Donnell fatigue design curve already includes a safety factor of 2 on the stress or a safety factor of 20 on the number of cycles as recommended by SRP Section 4.2 (1)(A)(ii). Therefore, the staff finds this criterion to be acceptable. Evaluation In Section 3.2.3 of APR1400-F-M-TR-13001 , Revision 1, the applicant present ed the results of a fatigue analysis for the PLUS7 fuel assembly.

The staff noted that TCD (which is not included in FATES3B) has a significant impact on the prediction of cumulative fatigue damage fraction. The staff issue d RAI 7954, Question 16 , requesting the applicant to describe how the fatigue analysis would be performed on a cycle

-specific basis given the demonstration that FATES3B without TCD is inadequate to assess the fatigue damage fraction. In its response to RAI 7954, Question 16 (ML17223B382), the applicant clarified that this methodology will not be used on a cycle

-specific basis

, but rather will demonstrate that the PLUS7 fuel design can meet fatigue limits within the requested operational window. In Section 3.4.3 of APR1400-F-M-T R-13001 , Revision 1, the applicant addressed burnup TCD by performing the fatigue analysis with the same modified version of FATES3B used in the cladding strain analysis (which replaced the Lyons thermal conductivity model with the modified NFI thermal conductivity model). The revised analysis also included a more limiting radial peaking factor. The resulting total cumulative fatigue damage factor remained below the limit. Based on the staff's review of the revised cladding fatigue analysis, the staff finds that the PLUS7 fuel assembly complies with the cladding fatigue criterion for the requested operating limits. Therefore, the staff finds this to be acceptable.

4.2.1.4 Cladding Oxidation and Hydriding Basis In Section 3.2.

4 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that the , "[f]uel system will not be damaged due to excessive oxidation under normal operation including AOOs." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding fuel system damage and fuel rod failure. Therefore , the staff finds this basis to be acceptable.

Criteria In Section 3.2.

4 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d , "[t]he best estimate cladding oxide thickness shall be less than

[

]." It is further state d that , "[t]he cla d hydrogen pickup is limited to

[

] at EOL to preclude loss of ductility due to hydrogen embrittlement by formation of zirconium hydride pellets.

" The staff notes that these criteria have been widely used by the industry and is consistent with the guidance provided in SRP Section 4.2 (1)(A)(iv) and (1)(B)(i). Therefore

, the staff finds these criteria to be acceptable.

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- PROPRIETARY INFORMATION Evaluation In Section 3.2.3 of APR1400-F-M-TR-13001 , Revision 1, the applicant present ed an end of life oxidation and hydrogen content evaluation of PLUS7 fuel using the PAD4 code.

The staff issued RAI 7954, Question 17 (ML15169A118)

, requesting the applicant to justify the use of a [

] hydrogen pickup fraction for ZIRLO. In its respon se to RAI 7954, Question 17 (ML15202A676), the applicant responded by comparing data and changing the pickup fraction to

[

] based on qualitative comparisons to data. The staff reviewed the applicant's response and performed a confirmatory analysis which supported the proposed [

] pickup fraction.

In Section 3.3.4 of APR1400

-F-M-TR-P , Revision 1, the applicant discusse d the use of the PAD code for the PLUS fuel design and concludes that a corrosion multiplier on the Zircaloy

-4 model of [ ] should be used to determine a best estimate oxide thickness for ZIRLO cladding. PAD4 typically uses a corrosion multiplier of [ ]. Based on KNF operational experience, a corrosion multiplier of [ ]

was selected for the analysis of PLUS7 fuel. The staff performed comparisons of the model and data and found that this is acceptable.

The applicant's analyses using the PAD4 code with the pickup fraction of [

] and the corrosion multiplier of [ ] resulted in oxidation and hydrogen content below the stated limits

. Based on the staff's review of the methods and analysis, the staff finds that the PLUS7 fuel design complies with the cladding oxidation criteria within the proposed operating limits.

Therefore, the staff finds this to be acceptable.

4.2.1.5 Fuel Rod Internal Pressure Basis In Section 3.2.5 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that the , "[f]uel system will not be damaged due to excessive rod internal pressure under normal operation

." This design basis is very similar to that given in SRP Section 4.2 (1)(A)(vi). Therefore, the staff finds this basis to be acceptable

. Criteria In Section 3.2.

5 of APR1400-F-M-TR-13001 , Revision 1, the applicant provide d three criteria for fuel rod internal pressure

(1) that cladding creep rate does not exceed fuel swelling rate (no cladding liftoff), (2) that reorientation of hydrides in the radial direction in the cladding shall not occur, and (3) that the radiological do se consequences of departure from nucleate boiling (DNB) failures are within the specified limits.

These criteria are consistent with the guidance provided in SRP Section 4.2 (1)(A)(vi). Therefore, the staff finds the s e criteria to be acceptable.

Evaluation The applicant provide d a rod internal pressure evaluation in Section 3.2.5 of the TR based on the use of the FATES3B code to calculate the critical pressure where cladding creep rate exceeds the fuel swelling rate. The staff issued RAI 7954, Question 15 , requesting the applicant to provide the rod internal pressure limit used for the PLUS7 fuel and the basis for the OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION limit. In its response to RAI 7954, Question 15 (ML15202A673), (Reference 3

), the applicant stated that FATES3B calculates the critical pressure limit when the rod internal pressure exceeds the systems pressure.

The applicant also state d that this critical pressure is around

[ ]. The critical pressure is based on the cladding creep rate and the fuel swelling rate.

The applicant confirmed that FATES3B will use the PAD4 ZIRLO creep model which the staff has previously found to be acceptable (ML003735452) (Reference 13). The staff notes that the fuel swelling mode in FATES3B is known to overpredict swelling in comparison to results from the staff's confirmatory code, FRAPCON

[ ]. This will lead to overpredicting (non

-conservative) the lift

-off pressure. However, in this case

, the applicant demonstrated that the maximum rod internal pressure for the PLUS7 fuel within the operating envelope requested will not exceed the system pressure. Therefore, the staff finds that although FATES3B may overpredict the critical lift

-off pressure, for this application and the requested operating envelope, the rod internal pressure will not exceed system pressure

. This is consistent with guidance provided in SRP Section 4.2 (1)(A)(vi). The staff requested , in RAI 7954, Question 18 (ML15169A118)

, for the applicant to provide a sample calculation showing rod internal pressure for a bounding power history up to a rod

-average burnup of 62 GWd/MTU. In its response to RAI 7954, Question 18 (ML15202A675)

, as supplemented by letter dated August 11, 2017 (ML17223B382), the applicant provided the information requested. The staff performed confirmatory calculations using FRAPCON, which are shown in Figure 3. It can be seen that the FATES3B calculation of rod internal pressure predicts the same or greater rod internal pressure as those calculated using FRAPCON.

In Section 3.4.4 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that TCD is inherently accounted for in FATES3B, since the model was calibrated to measure data for a full range of fuel rod burnup and operating conditions. There was insufficient information supporting this claim for the staff to review. I n its response to RAI 7954, Question 18, the applicant additionally discusse d conservatisms that are included in the original FATES3B calibration process , and in the design methodology

, which the applicant described would offset any potential increase of rod internal pressure due to reduction of void volume caused by TCD

-induced fuel thermal expansion. The staff's confirmatory analysis based on the applicant's response to RAI 7954, Question 18 , confirms that the applicant's results are conservative in comparison to FRAPCON.

OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION Figure 3: Comparison of the applicant and FRAPCON calculated rod internal pressure for sample calculation.

The applicant evaluated DNB propagation using the NRC approved methodology based on the INTEG code. This analysis was based on limiting DNB transients and internal pressure. The results from the applicant's analysis demonstrate d that the calculated strains would not induce DNB propagation. The staff has reviewed the information provided and finds that the PLUS7 fuel assembly design would not exhibit DNB propagation within the requested operating parameters.

The applicant evaluated the PLUS7 fuel assembly in terms of hydrogen reorientation by noting that the calculated rod internal pressure is well below the pressure at which hydrogen reorientation occurs.

Based on the applicant's analysis, as supported by the staff's confirmatory analysis, the staff finds that the PLUS7 fuel assembly meets the fuel rod internal pressure criteria for the requested operating parameters. The staff notes that since the calculated maximum rod internal pressures for the requested operating parameters are below system pressure, the DNB propagation and hydrogen reorientation criteria are not necessary per the guidance provided in SRP Section 4.2

they do, however, provide additional defense in depth. 4.2.1.6 Internal Hydriding Basis In Section 3.2.6 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that the , "[f]uel system will not be damaged due to excessive hydriding

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding hydriding and fuel rod failure

. Therefore, the staff finds this basis to be acceptable.

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- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION Criteria In Section 3.2.6 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that to avoid internal hydriding the moisture content in the pellet will remain below 2.0 ppm.

The staff has reviewed the information provided and notes that this criterion is more conservative tha n the guidance provided in SRP Section 4.2 (1)(B)(i). Therefore, the staff finds this criterion to be acceptable.

Evaluation The applicant provided a qualitative analysis which discusses the manufacturing process controls as the method by which the internal hydriding analysis is met.

The staff reviewed the applicant's evaluation and finds that the PLUS7 fuel assembly internal hydrogen levels are kept lower than the limits provided by the guidance in SRP Section 4.2(1)(B)(i). Therefore, the staff finds that the PLUS7 fuel assembly complies with the guidance provided in SRP 4.2(1)(B)(i) regarding internal hydriding

. 4.2.1.7 Cladding Collapse Basis In Section 3.2.7 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[f]uel rod failure will not occur due to cladding collapse

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding cladding collapse and fuel rod failure. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 3.2.7 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that radial buckling of the clad will not occur if axial gaps longer than

[

] will not occur between fuel pellets and that the plenum spring radial support capacity is sufficient to prevent clad collapse. Therefore, the staff finds th is criteri on to be acceptable.

Evaluation In Section 3.2.

7 of APR1400-F-M-TR-13001 , Revision 1, the applicant present ed a clad collapse analysis based on a previously approved methodology, CENPD

-187-P-A (Reference 11), which uses the CEPNAFL and FATES3B codes.

The applicant also state d that the PAD4 ZIRLO creep model is used as part of this evaluation.

The applicant provided results from its cladding collapse evaluation which conclude that cladding collapse will not occur within the operating range proposed over the entire life of the rods for both UO 2 and Gd 2 O 3-UO 2 rods. The results also confirmed that the plenum spring provides sufficient radial support to the cladding to preclude cladding collapse.

The applicant additionally stated that operational experience indicate s that collapse does not occur for fuel rods with initial pellet density of 95 percent theoretical density and having initial helium pressurization of [

].

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- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION The staff finds that by using approved codes and methods, the applicant has demonstrated that the clad collapse criterion is met and the fuel cladding in the PLUS7 fuel assembly will not experience buckling.

4.2.1.8 Overheating of Cladding Basis In Section 3.2.8 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[f]uel rod failure will not occur due to the overheating of cladding under normal operation including AOOs." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2

, regarding thermal margin and fuel rod failure

. Therefore, the staff finds this basis to be acceptable.

Criteria In Section 3.2.8 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[t]here should be a 95% probability at the 95% confidence level that a hot fuel rod in the reactor core will not experience a DNB during normal operation or AOOs. For postulated accidents, the rods that experience DNB are assumed to fail for radiological dose calculation purposes.

" SRP Section 4.2 (1)(B)(iii) states that ensuring a thermal margin such as DNBR for PWRs is acceptable to demonstrate cladding will not overheat and that for postulated accidents, the rods that experience DNB should be assumed to fail for radiological dose calculation purposes. SRP Section 4.4 states that there should be at least a 95

-percent probability at the 95

-percent confidence level that the hot fuel rod in the core does not experience a DNB or transition condition during normal operation or AOOs.

The staff finds that this criterion is consistent with the guidance provided in SRP Sections 4.2 and 4.4. Therefore, the staff finds this criterion to be acceptable.

Evaluation The evaluation for overheating cladding is addressed in the plant specific transient and acciden t analysis (Chapter 15 of DCD). Therefore, the applicant did not request review and approval for the evaluation itself as part of this TR. 4.2.1.9 Overheating of Fuel Pellets Basis In Section 3.

2.9 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that , "[f]uel rod failure will not occur due to the overheating of fuel pellets under normal operation including AOOs. For postulated accidents, the total number of rods that experience centerline melting should be considered for radiological dose calculation

." The staff has reviewed the information provided and notes that this criterion is consistent with the guidance provided in SRP Section 4.2 regarding thermal margin and fuel rod failure. Therefore, the staff finds this basis to be acceptable.

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- PROPRIETARY INFORMATION Criteria In Section 3.2.9 of APR1400-F-M-TR-13001 , Revision 1, the applicant state d that during normal operation and AOOs the fuel melting temperature will not be exceeded. The melting temperature of unirradiated UO 2 is 5080 °F and decreases by

[ ] per 10 GWd/MTU and [ ] for each weight percent of Gd 2 O 3. During a postulated accident if the fuel melts

, the fuel rod is considered to be failed and will be considered in the radiological consequence calculation.

The staff notes that the applicant's criteria is consistent with the high level guidance provided in SRP Section 4.2 (1)(B)(iv). However, in order to independently confirm the applicant's specific criteria, the staff compared th e applicant's correlation for fuel melting temperature versus the corresponding correlation in FRAPCON and found it to be very conservative. Figure 4 shows the results of this comparison. Based on the discussion provided by the applicant, and with supporting staff confirmatory analysis, the staff finds this criteri a to be acceptable. Figure 4: Comparison of the applicant and FRAPCON fuel melting limits for UO2 and Gd 2 O 3-UO 2 with 10 wt

. percent Gd 2 O 3. Evaluation In Section 3.2.9 of APR1400-F-M-TR-13001 , Revision 1, the applicant use d FATES3B and the previously approved C-E Fuel Evaluation Model methodology to calculate the maximum fuel centerline temperature. FATES3B is used to calculate a power

-to-melt curve as a function of burnup and the anticipated maximum powers are compared to this curve.

The staff notes that FATES3B will underpredict fuel centerline temperature as a function of burnup because TCD is not considered in FATES3B. The staff issued RAI 7954, Question 12

OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION (ML15169A118)

, requesting the applicant to discuss how an analysis will be performed for UO 2 and Gd 2 O 3-UO 2 fuel each cycle to ensure that the fuel will not melt since FATES3B is non

-conservative in its prediction of fuel centerline temperature. In its supplemental response to RAI 7954, Question 12

, dated August 11, 2017 (ML17223B382), the final methodology was eventually changed to use FATES3B and a temperature penalty on fuel centerline temperature as presented in Section 3.4.5 of APR1400-F-M-TR-13001 , Revision 1. Using FATES3B and temperature penalty, the applicant produced a power-to melt curve for UO 2 and Gd 2 O 3-UO 2 fuel. These power

-to-melt curves are compared to the radial fall

-off curves from Figure 3-1 of APR1400-F-M-TR-13001, Revision 1, and are confirmed to be greater than the maximum requested power for the entire burnup range.

These comparisons are shown in Figure 3-3 of the TR. The staff performed confirmatory calculations to confirm that the power to melt curves produced using FATES3B with the temperature penalty were acceptable. Figure 5 shows these comparisons with FRAPCON in blue and FATES3B with the temperature penalty in red. It can be seen in this figure that FATES3B with the temperature penalty predicts a lower power to melt than FRAPCON for all burnup ranges and both fuel types. This demonstrates that FATES3B with the temperature penalty is conservative relative to FRAPCON.

Figure 5: Comparison of the applicant and FRAPCON calculated power to melt for sample calculations with UO 2 and Gd 2 O 3-UO 2 rods. Based on the applicant's analysis, as independently confirmed by the staff with FRAPCON, the staff finds that the PLUS7 fuel design will meet the overheating of the fuel pellets criteria within the operating limits proposed.

Therefore, the staff finds this to be acceptable.

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- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION 4.2.1.10 Excessive Fuel Enthalpy Section 3.2.10 of APR1400-F-M-TR-13001, Revision 1 , states that the fuel rod failure is not underestimated for postulated accidents. Additionally, Section 3.2.10 of the TR provides failure criteria for determining the number of failed fuel rods. The staff evaluated these criteria and find s that these criteria match the criteria for determining the number of fuel failure s provided in Appendix B to SRP 4.2 (Reference 6) and are therefore

, acceptable. The TR, however, states that the evaluation of excessive fuel enthalpy is addressed in the plant specific transient and accident analysis. Accordingly, no finding regarding the evaluation of this criterion is made in this SER.

4.2.1.11 Pellet-to-Cladding Interaction Section 3.2.11 of APR1400-F-M-TR-13001, Revision 1 , states that fuel rod failure will not occur due to pellet cladding interaction (PCI) under normal operation including anticipated operational occurrences. The TR provided criteria on cladding strain and fuel melting. The staff finds th i s criteria to be consistent with the acceptance criteria provided in SRP Section 4.2, Item II.1.B.vi (Reference 6). The staff's evaluation of the cladding strain and fuel melt analyses are provide d in Section 4.2.1.2 and Section 4.2.1.9 of this SER, respectively.

4.2.1.12 Bursting Section 3.2.12 of APR1400-F-M-TR-13001, Revision 1 , states that fuel rod failure s are permitted during postulated accidents but that they will be accounted for in the dose analysis. The staff determined that the criteria provided in the TR matches the acceptance criteria provided in SRP Section 4.2, Item II.1.B.vii (Reference 6). Therefore, the staff finds this to be acceptable. The TR, however, states that the evaluation of th is criteri on is addressed in the plant specific transient and accident analysis. Accordingly, no finding regarding the evaluation of this criterion is made in this SER.

4.2.1.13 Cladding Embrittlement Section 3.2.13 of APR1400-F-M-TR-13001 , Revision 1 , states that coolability is always maintained under postulated accidents. The staff evaluated the criteria provide d in the TR and find them to match the requirements of 10 CFR 50.46(b)(1

), "Peak cladding temperature

," and 10 CFR 50.46(b)(2)

, "Maximum cladding oxidation

." Therefore, the staff finds this to be acceptable. The TR, however, states that the evaluation of this criteri on is addressed in the plant specific transient and accident analysis. Accordingly, no finding regarding the evaluation of this criteri on is made in this SER.

4.2.1.14 Violent Expulsion of Fuel Section 3.2.14 of APR1400-F-M-TR-13001, Revision 1 , states that coolability is always maintained under postulated accidents. The staff evaluated the criteria provide d by the applicant and find s them to match the core coolability criteria provided in Appendix B to SRP 4.2 (Reference 6). Therefore, the staff finds this to be acceptable. Additionally, the staff compared these criteria to the criteria provided in Draft Regulatory Guide DG-1327 (Reference 8) and find s the criteria specified in Section 3.2.14 of the TR to be more restrictive than the criteria provided in DG-1327. The TR, however, states that the evaluation of this c riteri on is addressed in the plant specific transient and accident analysis. Accordingly, no finding regarding the evaluation of this criteri on is made in this SER.

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- PROPRIETARY INFORMATION 4.2.1.1 5 Generalized Cladding Melting Section 3.2.16 of APR1400-F-M-TR-13001, Revision 1 , states that cladding embrittlement is more stringent than the melting criteria provided in Section 3.2.13 of the TR. The staff compared the 1204 °C (2200 °F) embrittlement criteria to the solidus temperature for Zircaloy provided in MATPRO (Reference

9) and determined that the embrittlement temperature is much lower than the temperature at which zirconium based alloys begin to melt. Accordingly, the staff finds that the cladding embrittlement criterion is more stringent than the generalized cladding melting criterion for ZIRLO.

4.2.1.1 6 Fuel Rod Ballooning Section 3.2.16 of the TR states that coolability is always maintained under postulated accidents and that fuel rod ballooning must be accounted for in the analysis of the core flow distribution. The staff finds that this basis is consistent with the acceptance criterion provided in SRP Section 4.2, Item II.1.C.iv (Reference 6). The TR, however, states that the evaluation of this criterion is taken into account in the plant specific LOCA analyses. Accordingly, no finding regarding the evaluation of this criterion is made in this SER.

5.0 LIMITATIONS AND CONDITIONS This TR does not address the evaluation of the fuel assembly for seismic and LOCA loads. As mentioned in Section 2.2.2.1 of the TR, the evaluation of fuel assembly for seismic and LOCA loads will be addressed in APR1400 DCD, Tier 2, Section 4.2.

The staff's approval is limited to a maximum peak rod average burnup of 60 GWD/MTU.

The fuel centerline temperature penalty presented in Section 3.4 of the TR is based on the operating parameters (e.g. peak linear heat generation rate) covered by the TR. The temperature penalty would need to be reevaluated for any core designs not bounded by these operating parameters.

The staff's approval is limited to baseload operations. Any applicant or licensee referencing this topical report who wishes to use non

-baseload operations must justify the fission gas release model for the intended operation and demonstrate that any impacted safety analyses do not exceed the limits.

6.0 CONCLUSION

The staff concludes that the PLUS7 fuel assembly has been designed so that

(1) the fuel system will not be damaged as a result of normal operation and anticipated operational occurrences, and (2) fuel damage during postulated accidents will not be severe enough to prevent control rod insertion when it is required, thereby meeting the related requirements of GDC 10, "Reactor design," GDC 27, "Combined reactivity control systems capability,"

and GDC 35, "Emergency core cooling,"

in Appendix A to 10 CFR Part 50 and 10 CFR 50.34. This conclusion is based on the following:

1. The applicant provided sufficient evidence that these design objectives will be met based on operating experience, prototype testing, and analytical predictions.

2. The applicant provided for testing and inspection of new fuel to ensure that it is within design tolerances at the time of core loading. The applicant made a commitment to OFFICIAL USE ONLY

- PROPRIETARY INFORMATION OFFICIAL USE ONLY

- PROPRIETARY INFORMATION perform online fuel failure monitoring and postirradiation surveillance to detect anomalies or confirm that the fuel has performed as expected.

Core coolability is addressed within the TR in part, but since the analyses supporting coolability such as large break LOCA are contained in separate topical and/or technical reports, no final finding is made by the staff in this SER regarding coolability.

The staff's review of core coolability following a LOCA will be documented in the SER associated with the topical report APR1400-F-A-TR-12004-P, "Realistic Evaluation Methodology for Large

-Break LOCA of the APR1400." The staff concludes that the applicant described methods of adequately predicting fuel rod failures during postulated accidents so that radioactivity releases are not underestimated and thereby meets the related requirements of 10 CFR 50.34 (for new reactors).

7.0 REFERENCES

1. "Transmittal of Topical Report APR1400

-F-M-TR-13001-P/NP, Revision 0, 'PLUS 7 Fuel Design for the APR1400' for Safety Evaluation," MKD/NW 0028L, dated September 17, 2013 (ML13298A413).

2. "The applicant Response to RAI 4

-7542," MKD/NW

-0020L, dated June 26, 2014 (ML14177A219).

3. "Response to RAI 5

-7954 on Topical Report 'PLUS7 Fuel Design for the APR1400,' APR1400-F-M-TR-13001, Rev. 0)," MKD/NW 0055L, dated July 21, 2015 (ML15202A673).

4. "Revised Response to RAI TOP 6-8322," MKD/NW 0130L, dated July 31, 2017 (ML17212B078

). 5. "Transmittal of Topical Report APR1 400-F-M-TR-13001-P Revision 1, "PLUS7 Fuel Design for the APR1400," MKD/NW 0205-L, dated August 11, 2017 (ML17223B416).

6. NUREG-0800, Standard Review Plan, Section 4.2, "Fuel System Design," Revision 3, dated March 2007.

7. WCAP-16500-P-A, "CE 16x16 Next Generation Fuel Core Reference Report," (ML072500331).

8. DG-1327, "Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents," dated November 2016 (ML16124A200).

9. NUREG/CR-6150, Volume 4, Revision 2, "SCDAP/RELAP5/MOD 3.3 Code Manual: MATPRO - A Library of Materials Properties for Light

-Water-Reactor Accident Analysis," dated January 2001 (ML010330400).

10. CENPD-404-P-A, "Implementation of ZIRLO TM Cladding Material in CE Nuclear Power Fuel Assembly Designs," dated November 2001 (ML031080082).

11. CENPD-187-P-A, Supplement 1

-P-A, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding" dated April 1985.

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- PROPRIETARY INFORMATION

12. CENPD-178-P, Rev. 1-P, "Structural Analysis of Fuel Assemblies for Seismic and Loss of Coolant Accident Loading

," August 1981.

13. WCAP-15063-P-A, Revision 1, "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," dated April 2000 (ML003735452).