The NRC issued a final safety evaluation report (FSER) for the ESBWR design in March 2011. The FSER provides the basis for issuance of a design certification under Subpart B to 10 CFR part 52 and a final design approval under Subpart E to 10 CFR part 52. The GEH has requested the NRC provide its design approval for the ESBWR design under Subpart E. The final design approval for the ESBWR design will be issued before publication of a final rule.
The significant technical issues that were resolved during the review of the ESBWR design are the regulatory treatment of non-safety systems (RTNSS), containment performance, control room cooling, steam dryer methodology, feedwater temperature (FWT) domain, aircraft impact assessment and the use of Code Case Nâ€“ 782.
Regulatory Treatment of Non-Safety Systems
The ESBWR relies on passive systems to perform safety functions credited in the design basis for 72 hours following an initiating event. After 72 hours, non- safety systems, either passive or active, replenish the passive systems in order to keep them operating or perform post- accident recovery functions directly. The ESBWR design also uses nonsafety- related active systems to provide defense-in-depth capabilities for key safety functions provided by passive systems. The challenge during the review was to identify the non-safety systems, structures and components (SSCs) that should receive enhanced regulatory treatment and to identify the appropriate regulatory treatment to be applied to these SSCs.
Such SSCs are denoted as â€˜â€˜RTNS SSCs.â€™â€™ As a result of the NRCâ€™s review, the applicant added Appendix 19A to the DCD to identify the nonsafety systems that perform these post-72 hour or defense-in-depth functions and the basis for their selection. The applicantâ€™s selection process was based on the guidance in SECYâ€“94â€“084, â€˜â€˜Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs.â€™â€™
To provide reasonable assurance that RTNSS SSCs will be available if called upon to function, the applicant established availability controls in DCD Tier 2, Appendix 19ACM, and Technical Specifications (TS) in DCD Tier 2, Chapter 16, when required by 10 CFR 50.36. The applicant also included all RTNSS SSCs in the reliability assurance program described in Chapter 17 of DCD Tier 2 and applied augmented design standards as described in DCD Tier 2, Section 19A.8.3. The NRC finds the applicantâ€™s implementation of the RTNSS process described in the DCD acceptable.
The passive containment cooling system (PCCS) maintains the containment within its design pressure and temperature limits for design-basis accidents. The system is passive and does not rely upon moving components or external power for initiation or operation for 72 hours following a loss- of-coolant accident (LOCA). The PCCS and its design basis are described in detail in Section 6.2.2 of the DCD Tier 2. The NRC identified a concern regarding the PCCS long-term cooling capability for the period from 72 hours to 30 days following a LOCA. To address this concern, the applicant proposed additional design features credited after 72 hours to reduce the long-term containment pressure. The features are the PCCS vent fans and passive autocatalytic recombiners as described in DCD Tier 2, Section 6.2.1. These SSCs have been indentified in DCD Appendix 19A as RTNSS SSCs.
The applicant provided calculation results to demonstrate that the long-term containment pressure would be acceptable and that the design complies with general design criterion (GDC) 38. The NRCâ€™s independent calculations confirmed the applicantâ€™s conclusion and the NRC accepts the proposed design and licensing basis. The NRC also raised a concern regarding the potential accumulation of high concentrations of hydrogen and oxygen in the PCCS and isolation condenser system (ICS), which could lead to combustion following a LOCA. The applicant modified the design of the
PCCS and ICS heat exchangers to withstand potential hydrogen detonations. The NRC concludes that the design changes to the PCCS and ICS are acceptable and meet the applicable requirements.
Control Room Cooling
The ESBWR primarily relies on the mass and structure of the control building to maintain acceptable temperatures for human and equipment performance for up to 72 hours on loss of normal cooling. The NRC had not previously approved this approach for maintaining acceptable temperatures in the control building. The applicant proposed acceptance criteria for the evaluation of the control building structureâ€™s thermal performance based on industry and NRC guidelines. The applicant incorporates by reference an analysis of the control building structureâ€™s thermal performance as described in Tier 2, Sections 3H, 6.4, and 9.4. The applicant also proposed ITAAC to confirm that an updated analysis of the as-built structure continues to meet the thermal performance acceptance criteria. The NRC finds that the applicantâ€™s acceptance criteria are consistent with the advanced light-water reactor control room envelope atmosphere temperature limits in NUREGâ€“1242, â€˜â€˜NRC Review of Electric Power Research Instituteâ€™s Advanced Light Water Reactor Utility Requirements Document,â€™â€™ and the use of the wet bulb globe temperature index in evaluation of heat stress conditions as described in NUREGâ€“0700, â€˜â€˜Human- System Interface Design Review Guidelines.â€™â€™ The NRC finds the control building structure thermal performance analysis and ITAAC acceptable based on the analysis using bounding environmental assumptions which will be confirmed by the ITAAC. Accordingly, the NRC finds that the acceptance criteria, control building structure thermal performance analysis, and the ITAAC, provide reasonable assurance that acceptable temperatures will be maintained in the control building for 72 hours. Therefore, the NRC finds that the control building design in regard to thermal performance conforms to the guidelines of Standard Review Plan Section 6.4 and complies with the requirements of the general design criteria of 10 CFR part 50, Appendix A, GDC 19.
Feedwater Temperature Operating Domain
In operating boiling-water reactors the recirculation pumps are used in combination with the control rods to control and maneuver reactor power
level during normal power operation. The ESBWR design is unique in that the core is cooled by natural circulation during normal operation, and there are no recirculation pumps. In Chapter 15 of the DCD, GEH references the licensing topical report (LTR) NEDOâ€“ 33338, Revision 1, â€˜â€˜ESBWR Feedwater Temperature Operating Domain Transient and Accident Analysis.â€™â€™ This LTR describes a broadening of the ESBWR operating domain, which allows for increased flexibility of operation by adjusting the FWT. This increased flexibility accommodates the so-called â€˜â€˜softâ€™â€™ operating practices, which reduce the duty (mechanical stress) to the fuel and minimize the probability of pellet- clad interactions and associated fuel failures.
By adjusting the FWT, the operator can control the reactor power level without control blade motion and with minimum impact on the fuel duty. Control blade maneuvering can also be performed at lower power levels.
To control the FWT, the ESBWR design includes a seventh feedwater heater with high-pressure steam. FWT is controlled by either manipulating the main steam flow to the No. 7 feedwater heater to increase FWT above the temperature normally provided by the feedwater heaters with turbine extraction steam (normal FWT) or by directing a portion of the feedwater flow around the high-pressure feedwater heaters to decrease FWT below the normal FWT. An increase in FWT decreases reactor power, and a decrease in FWT increases reactor power. The applicant provided analyses that demonstrated ample margin to acceptance criteria. The NRC concludes that the applicant has adequately accounted for the effects of the proposed FWT operating domain extension on the nuclear design. Further, the applicant has demonstrated that the fuel design limits will not be exceeded during normal or anticipated operational transients and that the effects of postulated transients and accidents will not impair the capability to cool the core. Based on this evaluation, the NRC concludes that the nuclear design of the fuel assemblies, control systems, and reactor core will continue to meet the applicable regulatory requirements.
Steam Dryer Design Methodology
As a result of reactor pressure vessel (RPV) steam dryer issues at operating BWRs, the NRC issued revised guidance concerning the evaluation of steam dryers. The guidance requested analysis to show that the dryer will maintain its structural integrity during plant operation in spite of or in the face of acoustic and hydrodynamic fluctuating pressure loads. This demonstration of RPV steam dryer structural integrity consists of three steps:
(1) Predict the fluctuating pressure loads on the dryer,
(2) Use these fluctuating pressure loads in a structural analysis to qualify the steam dryer design, and
(3) Implement a startup test program for confirming the steam dryer design analysis results during the initial plant power ascension testing.
The Plant Based Load Evaluation (PBLE) methodology is an analytical tool developed by GEH to predict fluctuating pressure loads on the steam dryer. Section 3.9.5 of the DCD references the GEH LTR NEDEâ€“33313P, â€˜â€˜ESBWR Steam Dryer Structural Evaluation,â€™â€™ which references LTR NEDEâ€“33312P, â€˜â€˜ESBWR Steam Dryer Acoustic Load Definition,â€™â€™ which references the PBLE load definition method. The PBLE method is described in LTR NEDCâ€“33408P, â€˜â€˜ESBWR Steam Dryer-Plant Base Load Evaluation Methodology.â€™â€™ This LTR provides the theoretical basis for determining the fluctuating loads on the ESBWR steam dryer, describes the PBLE analytical model, determines the biases and uncertainties of the PBLE formulation, and describes the application of the PBLE method to the evaluation of the ESBWR steam dryer.
The NRCâ€™s review of the PBLE methodology concludes that it is technically sound and provides a conservative analytical approach for definition of flow-induced acoustic pressure loading on the ESBWR steam dryer. The application of the PBLE load definition process together with the design criteria from the American Society of Mechanical Engineers (ASME) Code, Section III, Article NGâ€“ 3000 in combination with the proposed start up test program provide assurance of the structural integrity of the steam dryer. Implementation of the analytical, design, and testing methodology for the ESBWR steam dryer demonstrate conformance with the general design criteria of 10 CFR part 50, Appendix A, GDCs 1, 2, and 4.
Aircraft Impact Assessment
Under 10 CFR 50.150, which became effective on July 13, 2009, designers of new nuclear power reactors are required to perform an assessment of the effects on the designed facility of the impact of a large, commercial aircraft. An applicant for a new design certification rule is required to submit a description of the design features and functional capabilities identified as a result of the assessment (key design features) in its
DCD together with a description of how the identified design features and functional capabilities show that the acceptance criteria in 10 CFR 50.150(a)(1) are met.
To address the requirements of 10 CFR 50.150, GEH completed an assessment of the effects on the designed facility of the impact of a large, commercial aircraft. The GEH also added Appendix 19D to DCD Tier 2 to describe the design features and functional capabilities of the ESBWR identified as a result of the assessment that ensure the reactor core remains cooled and the spent fuel pool integrity is maintained.
The NRC finds that the applicant has performed an aircraft impact assessment using NRC-endorsed methodology that is reasonably formulated to identify design features and functional capabilities to show, with reduced use of operator action, that the acceptance criteria in 10 CFR 50.150(a)(1) are met. The NRC finds that the applicant adequately describes the key design features and functional capabilities credited to meet 10 CFR 50.150, including descriptions of how the key design features and functional capabilities show that the acceptance criteria in 10 CFR 50.150(a)(1) are met. Therefore, the NRC finds that the applicant meets the applicable requirements of 10 CFR 50.150(b).