Activities for 4S USNRC licensing
Kyoko Ishii a ,*,Hisato Matsumiya a ,Norihiko Handa b
a Advanced System Design and Engineering Dept.,Isogo Nuclear Engineering Center,Toshiba Corporation,8,Shinsugita-cho,Isogo-ku,Yokohama 235-8523,Japan b
Nuclear Energy Systems and Service Div.,Toshiba Corporation,8,Shinsugita-cho,Isogo-ku,Yokohama 235-8523,Japan
a r t i c l e i n f o
Article history:
Received 14October 2010Received in revised form 27May 2011
Accepted 30May 2011Keywords:4S
Sodium-cooled fast reactor Design approval NRC
a b s t r a c t
4S (Super-Safe,Small and Simple)is a small sized sodium-cooled fast reactor being developed for the electricity supply in remote areas,high-temperature steam supply more than 400 C,seawater desali-nation,and hydrogen production.The system design of power output of 10MWe (30MWt)has been completed.The main feature is that it does not have to be refueled for a long period (i.e.30years for 10MWe version),and enable the reactor closure sealed during plant operation.Furthermore,the small size of the reactor makes the reactor building suitable for below grade installing.These two features can provide resolutions for the issues relevant to safety,security,and safeguard,which become much more serious matter internationally these days.
4S is a pool-type reactor which contains the whole primary cooling system in a vessel.For the purpose of reducing the maintenance requirements with the reactor,(1)re flectors to compensate for fuel burn-up instead of control rods,(2)electromagnetic pump (EMP)which has no rotating parts,and (3)residual heat removal system by natural circulation and natural air draft are adopted.Therefore,exchange of the reactor components is not required during plant operation,in addition to no needs for refueling.
Toshiba has initiated the U.S.Nuclear Regulatory Commission (NRC)pre-application review of 10MWe version for the purpose of applying for design approval (DA).A series of public meetings with
NRC has been held four times,and five technical reports have been submitted to NRC in preparation for DA application.Topics discussed in these meetings included,plant design,metallic fuel,safety design philosophy,safety analysis,measures against severe accident,phenomena identi fication and ranking table (PIRT),etc.Some useful comments and questions on the issues regarding the speci fic feature of 4S as well as sodium-cooled fast reactor were raised by NRC at the public meetings.Among them,those items which are applicable to general sodium-cooled fast principal design criteria,guideline for safety analysis,validation and veri fication for safety analysis code,quality requirements,severe accident,and emergency planning are presented in this paper.
Ó2011Elsevier Ltd.All rights reserved.
1.Introduction
Super-Safe,Small and Simple (4S)is a small sized sodium-cooled fast reactor being developed for the electricity supply in
remote areas,high-temperature steam supply more than 400 C,seawater desalination,and hydrogen production.The 30MWt version of 4S does not have to be refueled for 30years,and main-tain operati
on with reactor closure sealed,which make the control of radioactive materials easier and prevent theft of nuclear fuels.Furthermore,the small size of the reactor makes the reactor building suitable for installing below ground,which mitigates the effect of an aircraft impact.Thus,these two features stated above provide an answer to the issues including safety,security and safeguards e sometimes referred to as the “3S ”concept.
The United States Nuclear Regulatory Commission (NRC)pre-application review public meetings of 4S were held four times during 2007to 2008(Toshiba Corp.,2007;Toshiba Corp.,2008a;Toshiba Corp.,2008b;Toshiba Corp.,2008c ).Then,five technical reports other than the report on the plant description have been submitted to NRC.Items discussed in these meetings included,
Abbreviations:4S,Super-safe,small and simple;ATWS,anticipated transient without scram;AOO,anticipated operational occurrences;CDF,cumulative damage fraction;CRBR,Clinch River breeder reactor;DBA,design basis accident;EBR-II,experimental breeder reactor-II;EM Pump,electromagnetic pump;EPZ,emergency planning zone;FFTF,fast flux test facility;FMEA,fault mode effective analysis;GDC,general design criteria;IHX,intermediate heat exchanger;LBE,license base event;LWR,light water reactor;NRC,the United States Nuclear Regulatory Commission;PDC,principle design criteria;PIRT,phenomena identi fication ranking table;PRISM,pow
er reactor innovative small module;RVACS,reactor vessel auxiliary cooling system;SG,steam generator.
*Corresponding author.Tel.:þ81457702415;fax:þ81457702317.E-mail address:kyoko1.jp (K.
Ishii).Contents lists available at ScienceDirect
Progress in Nuclear Energy
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Progress in Nuclear Energy 53(2011)831e 834
plant design,metallic fuel,safety design philosophy,safety analysis, measures against severe accident,phenomena identification and ranking table(PIRT)(Wilson and Boyack,1998),etc.
The purpose of this report is to share the lessons we learned from the4S licensing activities that are applicable to general sodium-cooled fast principal design criteria,guidance for safety an
alysis,validation and verification for safety analysis code,severe accident,and emergency planning,some of which were raised in SECY-10-0034“Potential Policy,Licensing,and Key Technical Issues for Small Modular Nuclear Reactor Designs”(SECY-10-0034,2010).This paper includes the4S response to those issues, as well.
2.4S Overview
Fig.1shows the plant concept of4S that is referred to in this paper.4S is a pool-type reactor,so the whole primary cooling system is contained in a vessel.The reactor is situated in the containment vessel which consists of guard vessel and top dome. The heat from the primary system is transferred to the intermediate heat transport system through the intermediate heat exchanger (IHX).Then,steam is produced by the steam generator(SG) installed at intermediate system.
For the purpose of reducing the maintenance requirements with the reactor,following system and components are adopted:(1) Reflectors to compensate for fuel burn-up instead of control rods, (2)electromagnetic pumps(EM Pumps)which are immersed in sodium,and(3)reactor vessel auxiliary cooling system(RVACS) which removes residual heat from the reactor passively,using natural convection of air outside the guard vessel and the primary sodium in the reactor.Therefore,in additio
n to no needs for refu-eling,the reactor components require no exchange during opera-tion.A double wall heat transfer tube system is adopted for SG to prevent sodium water reaction.A single tube failure can be detected by tube failure monitoring system incorporated with SG.
In terms of licensing matters,the small size of reactor may have advantages comparing to the large scale reactors.For example,(1) the required demonstration tests for licensing can be conducted with small scale and cost,(2)the safety related systems and components can be reduced by adopting the passive safety systems, (3)the annual fee can be less in case a variable annual fee structure based on the thermal or electrical power limits of the reactor is approved by NRC,and(4)evaluation and measures against acci-dents of the spent fuel storage pool are not required because4S does not have it.For the decommissioning,4S is considered to follow the same decommissioning method as Fast Flux Test Facility (FFTF).
3.Development of principal design criteria
NRC has conducted two reviews of sodium-cooled fast reactors so Clinch River Breeder Reactor(CRBR)final safety evalua-tion report(NUREG-0968,1983)and Power Reactor Innovative Small Module(PRISM)preliminary safety evaluation report (NUREG-1368,1994).The design criteria
which NRC referred to for those review were Appendix A to10CFR50“General Design Criteria (GDC)for Nuclear Power Plants”and ANSI/ANS54.1(ANSI/ANS-54.1,1989).In preparation for pre-application review,“4S prin-cipal design criteria(PDC)”was established on the basis of Appendix A to10CFR50,taking in consideration the characteristic features of the4S design,the trend of revision of regulations for current light water reactors(LWRs),and ANSI/ANS54.1.The4S PDC was used as a baseline for the4S design,and it is reported to NRC how the4S safety system meets the criteria.In these processes,the deterministic methodology is used complemented by4S-specific risk insights currently available.
百变神龙Against this approach,NRC manifested their interest in knowing the process that the applicant used to establish PDC,since Toshiba made modification in its criteria from Appendix A to10CFR50for application to sodium-cooled fast reactors;ex.adding the requirement for protection against sodium reactions,changing the terminology of control rod to control element,etc.The issue NRC pointed out was the necessity to develop GDC for fast reactors by independent organization other than the applicant themselves. What the applicant requested is to show how they meet those GDC for sodium-cooled fast reactors.
Currently,the development of the regulatory framework for liquid metal sodium-cooled fast reactor is i
n progress under
ANS/
Fig.1.The plant concept of a4S.
K.Ishii et al./Progress in Nuclear Energy53(2011)831e834
832
ANSI54.1Standard Committee in cooperation with other interested parties including Toshiba.The conformance of the4S design to those criteria is to be demonstrated at design approval(DA)application.
4.Guidance for safety analysis
For the safety evaluation of4S,the license base event(LBE)and the safety acceptance criteria for sodium-cooled fast reactor needed to be settled.In order to do that,Toshiba directly adopted the latest guide10referring to the methodology for selection of the LBE and for safety analysis of the LWRs.First,failure mode and effect analysis (FMEA)was performed to identify transient and accident candidates for4S.Then,those events were categorized into three groups as described in the anticipated operational occurrence(AOO), design basis accident(DBA),and anticipated transient without scram (ATWS).Furthermore,conservative analysis conditions were estab-lished for reactivity temperature coefficients,material property values,and plant design data.This approach is consistent with the approach NRC expects applicant to use as stated in subsection3.4of SECY-10-0
deterministic engineering judgment com-plemented by insights from the PRA.The validity of the results of the event categorization will be discussed at DA application.
At the same time,safety acceptance criteria are needed to be established to evaluate the results of the safety analysis.The acceptance criteria for the safety analysis of the4S reactor are defined as follows,considering the guidance of SRP15.0(NUREG-0800,2007),and making allowance for differences between liquid metal reactors and LWRs;maintaining the integrity of fuel cladding and the core coolable geometry,the allowable radiation exposure at an exclusion area boundary,integrity of the primary coolant boundary,and the containment integrity.
Against this approach,a detail explanation pertaining to those safety acceptance criteria was requested to provide by NRC.Hence, Toshiba submitted a technical report(Toshiba Corp.,2009) explaining the4S safety acceptance criteria as follows.The prevention of core melt,maintaining the integrity of the fuel cladding and the core coolable geometry is defined as acceptance criteria for transients referring to the acceptance criteria developed for the Experimental Breeder Reactor-II(EBR-II)Mark-V metallic fuel pins.Cumulative Damage Fraction(CDF)is introduced for the criteria of cladding integrity.Then a criteria based on creep strain was introduced as the criteria for coolable geometry,because it was assumed that the integrity of coolable geometry is lost when the diameter o
f fuel cladding continues to expand due to creep effect until the gap between claddings,which is theflow pass of the coolant,is blocked.The validity of those criteria was remained to be future discussion after DA application.
For the design and demonstration of the fuel that has30years of life,the experience of EBR-II is referred.Although EBR-II fuel have been stored shorter period(1e3years)in the reactor,4S fuel has less burn-up rate,which is4.5%at maximum,compared to that of the EBR-II metallic fuel which is more than10to20%.Moreover,the amount of irradiation of the4S core(2Â1023n/cm2)is less than that of EBR-II(2e4Â1023n/cm2).Hence,it is possible to utilize the test data of EBR-II to evaluate the integrity of the4S fuel.
5.Validation and verification for safety analysis code
The safety performance of4S is analyzed with a plant dynamics code for a sodium-cooled fast reactor,the ARGO code(Tsuji et al., 2003),developed by Toshiba.Analysis codes and their validation for safety analysis were one of the NRC’s concerns from the beginning.NRC issued regulatory guide referring to validation and verification for the evaluation models that may be used to analyze transient and accident behavior.
During the public meetings,the need of validation and verifi-cation for the ARGO code was emphasized by NRC.Since the vali-dation and verification for the ARGO code is pursued by Toshiba in preparation for DA application,a report on this topic will be provided in the future.
The validation and verification of ARGO code are currently undergoing following the guidance issued by NRC(Regulatory Guide1.203,2005;NUREG-1737,2000).For the validation,the tests to demonstrate the applicability of evaluation models are identified using PIRT process,and those tests will be conducted in near future.The document summarizing the results of the validity of the analysis results for ARGO code will be submitted upon DA application.
6.Test for key safety components
In the past,Toshiba had conducted several tests for the reactor components related to the4S key components.The data obtained from those tests are utilized to substitute the demonstration tests for the4S components especially important to safety assurance.As indicated in Table1,those tests include the critical test using the cylindrical shaped reflector which surrounds the core,hydraulic test for advanced type subassembly with high fuel volume fraction, operating test of reflector drive mechanism which ascends upwards at very low speed withfine movement,radiation heat transfer tes
t to evaluate the performance of the residual heat removal from guard vessel,test of full scale EM Pump and double wall tube SG using sodium loop,and test of seismic isolator.
During the public meeting,an issue about quality of some data obtained by those tests stated above remained to be solved because the quality assurance program imposed by NRC has not strictly met. Hence,those data will be re-obtained in appropriate manner.
Currently,Toshiba is newly conducting some demonstration tests for advanced he sodium-immersed type EM Pump,which especially NRC paid attention to,which requires no replacement during plant lifetime;30years.The test of the EMP is currently ongoing using sodium test loop.For the SG,test to confirm inspection method of double wall tube during pre-service and in-service inspection is ongoing.
Table1
Test to support the4S design.
Design feature Verification item Required testing Status
Long cylindrical core with small diameter Reflector controlled core Nuclear design method of reflect
or controlled
core with metallic fuel
Critical experiment Done
High volume fraction metallic fuel core Confirmation of pressure drop in fuel subassembly Fuel hydraulic test Done
RVACS Heat transfer characteristic between vessel and air Heat transfer test of RVACS Done
EM pump Fabrication and design method
Stable characteristics
Sodium test of EM pump Done and ongoing
Steam generator(Double wall tubes) Welding method for tubes
Heat transfer characteristic
Leak detection Sodium test of steam generator
yangjiang
Leak detection test
Done and ongoing
平湖十八楼K.Ishii et al./Progress in Nuclear Energy53(2011)831e834833
7.Severe accident
One of the items seriously discussed was a design approach against core damage accidents.Prevention of core damage is the most important issue in the 4S safety design strategy.4S provides multiple design measures not to lead core damage under any conditions.Although these measures are not completely redun-dant,some of severe accident progressions are effectively blocked.At the first,all of 4S safety systems are designed based on requirements in 4S PDC which request redundancy of the active components in a safety system.Secondly,maximum use of passive or inherent features enhances safety,including metallic fuel with higher thermal conductivity which reduces accumulated enthalpy in a fuel slug,negative reactivity feedback temperature coef fi-cients,negative void reactivity,and shutdown decay heat removal by natural circulation and natural draft of air.Thirdly,evolutional design elements are introduced to focus mainly on prevention of core damage,including no refueling to minimize the possibility for entrainment of impurities,double wall tu
be SG with tube failure monitoring to prevent sodium water reaction,seismic isolation to withstand possible big earthquake,and backup core support structures to prevent core from falling by failure of the core support structure.It was reported to NRC that these safety measures or functions were enough to exclude every event sequence that leads to core damage.A schematic of the approach is shown in Fig.2.
Against this discussion,much more comprehensive approach along with a series of their policy statements (73FR 60612)(73FR 60612,2008)which have been issued since 1985was requested by NRC to provide.In particular,recent policy statement was strengthened for review that requests containment integrity at an aircraft impact.Hence,Toshiba is planning to submit a technical report including design conformance to the policy statements on severe accidents (50FR 32138)(50FR 32138,1985)and the measure against the issues related to the severe accidents at previous fast reactors.The 4S aircraft impact assessment is planned to be done according to the aircraft impact rule for new LWRs (74FR 28112)(74FR 28112,2009).However,the applicability of this process will be discussed upon DA application because this rule may not be applicable to non-LWR designs.Toshiba will comply with the regulation for non-LWR in case newly issued.
A system design based on ‘Defense-in-Depth ’is the key against core damage as long as determinist
ic license framework is adapted.In addition,it may not be appropriate to apply PRA to the safety system incorporating passive or inherent design features,because there is a dif ficulty in developing the failure probability of those features by conventional mathematical method.Toshiba continu-ously pursue a set of the 4S design for prevention of core damage accidents.An issue how to cut the chain of over conservatism in deterministic approach may require further discussion.
8.Emergency planning
According to 10CFR 50.47(c)(2)(10CFR 50.47,2007),the size of emergency planning zones (EPZs)may be determined on a case-by-case basis for reactors with power level less than 250MW thermal like 4S.At the pre-review meeting,the process of developing the emergency preparedness requirements such as whether only beyond DBAs were accounted for referring to a licensing scheme with an operating reactor was discussed.The 4S EPZ is developed accounting for both DBAs and beyond DBAs.Although core damage accidents shall be evaluated in deter-mining the EPZs,the core damage accidents are hard to assume for 4S due to the inherent safety features.Therefore,a hypothet-ical 100%fuel failure is to be assumed as beyond DBAs for eval-uation in determining the EPZs.9.Conclusion
Toshiba has entered NRC pre-application review of 4S in prep-aration for DA application.In the review process,the safety issues pertain to general sodium-cooled fast reactors were discussed with NRC.In this paper,several topics among them and our responses in the near future were reported to share the lessons learned from 4S NRC licensing activities.The safety acceptance criteria against “coolable geometry ”may be still open issue that will be discussed at DA review process.Furthermore,it is great importance how to assure or show that the proposed design is enough to prevent core damage.References
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EM Pump with no rotating parts Double wall tube SG Seismic isolation Backup core support
No refueling for plant lifetime
4S Principal Design Criteria based on ANS-54.1
Metallic fuel
Negative reactivity feedback Negative void reactivity Natural circulation
Passive or Inherent Designs Evolutionary Designs
Safety Design according to Deterministic General Design Criteria
Fig.2.A schematic of the approach for measures against severe accidents.
K.Ishii et al./Progress in Nuclear Energy 53(2011)831e 834
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