ML20028G292
| ML20028G292 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 01/31/1983 |
| From: | Hanson J EG&G, INC. |
| To: | Allen C Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20028G285 | List: |
| References | |
| CON-FIN-A-6472 EGG-NTAP-6152, EGG-NTAP-6152-DRFT, NUDOCS 8302070598 | |
| Download: ML20028G292 (89) | |
Text
EGG-NTAP-6152 January,'1983 COMPARISCN OF CLINCH RIVER BREEDER REACTOR DESIGN BASES ACCIDENTS WITH THOSE FOR LIGHT WATER REACTORS AND LIQUID-METAL-COOLED FAST REACTORS s
J. E. Hanson Idaho National Engineering Laboratory Operated by the U.S. Department of Energy
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1 Prepared for the U. S. NUCLEAR REGULATORY COMMISSION Ur. der DOE Contract No. DE-AC07-761001570 0 Fin No. A6472 g b b E b idaho ! 8002070599 830110 POR ADOCK 05000507 A PDR
h EGcG,.. . ,- FORM EG&G 396 iRev C1421 INTERIM REPORT Accession No. Report No. EGG-NTAP-6152 s Ccntract Program or Project
Title:
. Case Review of CRBR Auxiliary Systems and Accident Analysis Fin No. A6472 Subject of this Document:
Review of Design Basis Accidents for the Clinch River Breeder Reactor Type of Document: Informal Report Author (s): J. E. Hanson . Data of Document: January, 1983 R:sponsible NRC Individual and NRC Office or Division: C. A. Allen, NRC/NRR This document was prepared primarily for preliminary or internal use. it has not received full review and approval. Since there may be substantive changes, this document should not be considered final. EG&G Idaho Inc. Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. Under DOE Contract No. DE-AC07-76 tD01570 NRC FIN No. A6472 . INTERIM REPORT
EGG-NTAP-6152 January 1983 COMPARISON OF CLINCH RIVEP, BREEDER REACTOR DESIGN BASIS ACCIDENTS WITH THOSE F0P. LIGHT WATER REACTORS ANG LIQUIC-META' -CCOLED FAS~i REACTORS I J. E. Hanson EG&G Idaho, Inc. Idaho Falls, Idaho 83415 ~ Prepared for the IJ.S. NUCLEAR REGULATORY COMMISSION Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6472
ABSTRACT Design basis acciocnts considered in the design of other U.S. fast reactors and in the safety and licensing review of light water reactors
' vere compared to those for the Clinch River Breeder Reactor. Based on this sonewhat limited review it appears that the spectrum of Clinch River Breeder Reactor design basis accidents is as well defined as possible given present judgement and past experience in the design, construction and operation of nuclear power reactors.
l A6472 - Comparison of Clinch River Breeder Reactor Design Basis Accidents With Those For Light Water Reactors And Liquid-Metal-Cooled Fast Reactors ! l i
CONTENTS ABSTRACT.............................................................. i S U M MA R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i i i I. INTRODUCTION ..................................................... l . II. DISCUSSION ....................................................... 2 A. DESIGN BASIS 'CCIDENT DELINEATION ........................... 2 B. COMPARISrN OF CLINCH RIVER BREEDER REACTOR DESIGN BASIS ACCIDENTS WITH THOSE FOR LIGHT WATER REACTORS ... . . . . . . . 7 C. DESIGN BASIS ACCIDENTS UNIQUE TO CRBR ....................... 29 D. COMPARISON OF CLINCH RIVER BREEDER REACTOR DESIGN BASIS ACCIDENTS WITH THOSE FOR OTHER LIOUID-METAL-C OOL E D F A ST R E A C T O R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25
- 1. Fast F lux Test F acili ty ( FFTF ) . . . . . . . . . . . . . . . . . . . . . . . . . . 46 '
- 2. Experimental B reeder Reactor-II (EBR-II ) . . . . . . . . . . . . . . . . 55
- 3. Southwest Experimental Fast 0xide Reactor (SEFOR) ....... 65 4 LMFBR Conceptual Design Study--Large Developmental Plant (LDP) ............................................. 72
- 5. Enrico Fermi Atomic Power P l ant (FERMI-1 ) . . . . . . . . . . . . . . . 80 III. CONCLUSIONS ...................................................... 89 4
TABLES i Table 1. Summary of Comparison of CRBR PSAR Chapter 15 Design ; Basis Accident Events with Those for LWRs as Given in NUREG-0800, Standard Review Plan ......................... 8 L Table 2. Comparison of FFTF DBAs with Those for CRBR ................. 48 ' Table 3. Comparison of EBR-II DBAs with Those for CRBR ............... 57 r Table 4 Comparison of SEFOR DBAs with Those for CRBR ................ 66 ; i Table 5. Comparison of LDP DBAs with Those for CRBR .................. 73 , Table 6. Comparison of FERMI-I DBAs with Those for CRBR .............. 82 ii ,
SUMMARY
4 Design basis accidents considered in the design of the Experimental , Breeder Reactor-II (EBR-II), the Southwest Experimental Fast Reactor 4 (SEFOR), the Enrico Fermi Atomic Power Plant (FERMI-I), the Fast Flux Test Facility (FFTF) and the Large Developmental Plant (LDP) were compared with those presented in Chapter 15 of the Clinch River Breeder Reactor Preliminary Safety Analys'is Report. In general, DBA delineation for the older plants (EBR-II, FERMI-I, and SEFOR) did not appear as well developed as that for the more recent f ast reactor desigas, i.e., FFTF, CRBR and LDP. One or the crincipal reasons for this is that this present somewhat f' limiteci study was restricted to review of the accident analysis sections of the Hazaros Summary Reg. orts for these earlier reactors. In each case, the Hazards Summary Reports precedsd cevelopment of the Standard Format and l Content and Standard Review Plan forraats. Additionally, in the early plants, hypothetical core disrupt vei accidents and fuel and core stability received major attention as they were perceived to be the major fast reactor safety issues at that time. With the advent of the Standard Format j and Content and Standard Review Plan formats, the safety analysis reports for FFTF, CRBR and LDP are more consistent and present in one place ] i analyses of a more complete set of design basis accidents. Despite the i differences in format and the degree to which events in the balance of 1 l plant systems were or were nct discussed in the Hazards Summary Reports for I the earlier plants, it is clear that DBA delineation for CRBR reflects the ) experience which has been obtained from the design, construction and I operation of the earlier plants. Moreover, to the extent that it is l { applicable, light water reactor experience is also reflected in CRBR DBA delineation. ) iii i _ , .. . , ~ . . .-.- , m, .,, . - _ , . - , _ , . _ . .. _ ,_ . _ . . . - . , . - _ _ _ , , _ _ ,
COMPARISON OF CLINCH RIVER BREEDER REACTOR DESIGN BASIS ACCIDENTS WITH THOSE FOR LIGHT WATER REACTORS l AND LIOUID-METAL-COOLED FAST REACTORS l l i I. INTRODUCTION , To complement the construction permit review of the Clinch River , Breeder Reactor.(CRBR), design basis accidents (DBAs) used in other U.S. fast breeder reactor designs were reviewed as a means of assessing the completeness of those used in the CRSR design. This study also complemer.ts the analysis of design basis accident delineation to be performed by the applicants wherein the rationale and justification for their selectior, of CRBR DBAs were to be examined (the report is to be issued in December 1982). Additionally, a continuing study of DBA delineation for CRBR is , incorporated in the applicants probabilistic risk assessment for CRBR. This work, which is in progress as part of the applicants voluntary 4 compliance with the pertinent portion of NUREG-0718 will be completed prior to issuance of an operating license. Should the study reveal event
] sequences which have not received sufficient attention during the design of ! the plant they will be addressed prior to issuance of the operating license. A companion report covering DBA delineation for foreign fast reactors is being prepared by Dr. A. Agrawal, Brookhaven National Laboratory.
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i-II. DISCUSSION A. DESIGN BASIS ACCIDENT DELINEATION As is the case for light water reactors (LWRs), design basis events and accidents for liquid-metal-cooled fast breeder reactors (LMFBRs) have been derived in a deterministic manner, i.e., from past experience in the e design, construction and operation of plants such as EBR-I, EBR-II, SEFOR, FERMI-l and FFTF. To the extent that it is applicable, LWR experience is also drawn upon. To a somewhat lesser degree the experience available from foreign fast reactor orograms is also drawn upon to complement the development of accident delineation. It should be borne in mind, however, . that many of the foreign fast reactors are based on U.S. technology. 10 the success in those programs, most notably in France, is a reflectisa of the i effic acy of U.S. practice. In present practice, f ast reactor design bas,ir accident delineation parallels that for LWRs. The following excerpts from the Rosenthal and Check testimony
- are particularly enlightening and directly applicable to '
fast-reactor design basis accident delineation.
"The term design basis is used in two senses. First, the term is defined in 10 CFR 50.2(u): Design basis means that information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design.
These values may be (1) restraints derived from generally accepted ' state of the art' practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which a structure, system, or component must meet its functional goals."
"An explicit list is not provided in the regulations. One must extract such a list on a system-by-system basis from ,
the FSAR, plant Technical Specifications, applicable reference or topical reports, and related occuments." 1
- a. NRC Staff Testimony of Jack Rosenthal and Paul S. Check Relative to UCS i
Contention 13, In the Matter of Metropolitan Edison Company (Three Mile Island Nuclear Station Unit 1 Restart), Docket No. 50-289. l 2
"The term is also used to mean the whole of the standards, criteria, guidance, and performance specifications applicable to the facility."
"This includes: conformance with the General Design l Criteria, regulatory guidance (e.g., Regulatory Guides and the Standard Review Plan), quality assurance programs, industrial design and construction standards, and commercially oriented (warrantee) requirements which affect plant reliability. The design bases also include plant
. analyses of tne response of various systems to specific design-basis events."
" Design basis events are the set of prescribed anticipated
. operational occurrences and accidents used to assess the way specific systems respond to upset conditions. Operational occurrences are events or conditions expected to occur one i or more times during the life of a nuclear ooser unit; ' accidents are events expected to occur less frequently if at all."
" Design basis events provide analytic tests of the design. ;
These tests consist of sample challenges to the plant safety J systems; they are used to provide an analytical basis for deterT.ining if installed or proposed safety features can ' cope adequately with the postulated event. These tests have been chosen baseo on engineering studies and engineering judgment of specific systems features." .
" Potential radiological consequences of design-basis events are assessed to ensure that predicted consequences are less than the guidelines of 10 CFR 100. They are treated as
' credible' events for purposes of Part 100."
"The term ' Class 9' events is derived from a proposed rule published by the AEC in 1971. The proposed rule, which has ,
now been withdrawn by the NRC, set forth a system of t classification of potential accidents for use in Staff NEPA : assessments. It set forth a spectrum of accidents divided into nine classes rangir.g from the most trivial to the most severe for purposes of evaluating environmental risk." l l
" Class 8 events were characterized as '. . . those ;
considered in safety analysis reports and AEC staff safety : evaluations.' They are used, together with highly conservative assumptions, as the design-basis events to , establish the performance requirements of engineered safety l features. The highly conservative assumptions and { calculations used in AEC safety evaluations are not suitable
- for environmental risk evaluation, because their use would
- result in a substantial overestimate of the environmental
; risk. For this reason, Class 8 events shall be evaluated [,
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realistically. Consequences predicted in this way will be far less severe than those given for the same events in safety analysis reports where more conservative evaluations are used."
" Class 9 events were characterized as '. . . involv(ing) sequences of postulated successive failures more severe than those postulated for the design basis for protective systems .
and engineered safety features.' Their consequences could be severe. However, the probability of their occurrence is so small that their environmental risk is extremely low. . Defense in depth (multiple physical barriers), quality assurance for design, manufacture, and operation, continued surveillance and testing, and conservative design are all applied to provide and maintain the required high degree of assurance that potential accidents in this class are, and will remain, sufficiently remote in probability that the environmental risk is extremely low. for these reasons, it is not necessary to discuss such events in applicants' Environmental Reports."
" Breach of containment and core melt accidents have been designated Class 9 events. . These terms have been often used interchangeably with the term Class 9 but such use is imprecise. Class 9 events could have radiological consequences ranging from insignificant to severe. For example, core damage events not involving loss of ccntainment integrity would have fairly limited radiological consequences." "As indicated in the proposed rule, events beyond the design basis were considered by the Commission to have sufficiently low probability of occurrence, and hence environmental risk was considered to be extremely low. These events, characterized as beyond the design basis, were not explicitly assessed in determining the adequacy of the facility design. For purposes of Part 100 these events were considered as 'not credible'." "Over the course of years the staff has used several approaches to determine whether a particular sequence is properly characterized as credible. Reactor regulation has been a developing process, with new information and techniques being incorporated as they have become available and verified. But fundamentally, the staff uses engineering judgement informed by engineering assessment of the performance characteristics of the various systems and -
components in a nuclear power reactor and of the kinds of system or component failures that may occur. This is often called a mechanistic or deterministic approach. It relies - upon the composite of engineering experience and expertise of the staff, supplemented by the engineering experience and 4
expertise of the ACRS, and with substantial contribution f from the engineering experience and expertise of designers, builders, and operators of nuclear power reactors."
"In the early days of reactor safety assessments, in the early sixties, determining the kinds of accidents that a plant should be designed to cope with was a matter of group
' assessment of what kinds of things can go wrong. Effort was made to bound those events that might reasonably be expected to happen. This was accomplished principally by postulating
. tre failure of each major plant system in turn, and requiring that the plant manage the consequences, i.e., to ensure that predicted offsite doses remain within Part 100 guidelines."
"With progressively more sophisticated safety assessments being performed during the sixties and early seventies, this process finally led, by the mid-seventies, to the set of design-basis events (anticipated operational occurrences and accidents) now used by the staff in all case reviews to test the overall adequacy of design. These events were believed to represent a sound composite engineering judgement regarding the reasonable upper bounds for events which might occur. Also, they were thought to define a reasonable envelope of all credible events. Thus, plants were required to be designed to mitigate the consequences of those events which were considered ' credible'. Conversely, mitigative measures were not required for those events which were considered 'ir:redible'."
"However, more recently, with the increasing use of risk assessment and with the current perception of event sequences as a continuum of probabilities of occurrence with a concomitant continuum of consequences, the staff is extending its consideration of failure sequences."
"The staff now considers a spectrum of event sequences and employs a variety of ' fixes'. It is staff practice to
?ddress those event sequences designated as design-basis events primarily by requiring installation of emergency safety featrues (i.e., hardware), although the staff has also employed procedural measures to mitigate OBEs. Event sequences not designated as ' design-basis events' have been
' fixed' by a variety of means including the use of increased surveillance and testing of existing equipment or improving plant procedures and operator training, as well as some
, , hardware requirements. The goal of these ' fixes' is to I reduce the probability and/or the consequence of an event sequence. Selection of the means to implement one or more
' fixes' is based in part on risk assessment, but still predominantly or engineering judgment."
5 J
"However, we now explicity consider a much wider range of event sequences, some involving multiple failures and some involving systems not traditionally considered safety In this connection, we consider plant or system systems. j modifications which can eliminate the initiating event, or i can improve the capability of some other systems to I coalpensate for or cope with the initial malfunction, as well i as improvements in mitigating system characteristics. These ,
are all intended to assure that the likelihood of the candidate sequence is diminished to a low level relative to other potential reactor safety system malfunctions or that . the potential consequences of such an event is less severe than analyzed design basis events."
"The methods used to assess this wider range of sequences remains principally a deterministic assessment. Although we now sometimes use fault trees and event trees to help us to understand the sequence, step-by-step, the assessment of the adequacy of systems to safely terminate the event is still i based on the experience and judgement of the review st.aff involved."
" Quantitative probability assessment has been used to help identify areas of relative strength and weaknesses and to fo.us on areas which should be subject to engineering review. However, this procedure has not been used as the sole arbiter in the decision process; it is just one of several tools used in the process."
"The staff recognizes several reasons why quantitative probabilistic analysis is not the only procedure used:
(1) because there is a lack of sufficient failure-rate data and because of difficulties in establishing complete syrtem models, sound usessments which have been adequately tested are rare; (2) the staff does not have a numerical probability goal against which to assess co,ipliance; and (3) the staff believes its approach--which utilizes composite engineering experience and judgment--prou des a sound, comprehensive basis for our decisions."
"Nonetheless, even when the staff has emphasized detailed analysis of postulated events, the probability element has not been totally absent. The probability associated with calculated consequences was given limited (generally quantitative) consideration, and some element of probability was reflected, primarily in the selection of events or event sequences to be analysed. For example, the probability of .
two or more random equipment failures initiating an event was considered to be low enough to eliminate this situation from consideration as a design-basis event. (Failures of - the mitigating systems were considered in accordance with the single-failure criterion, Appendix A, 10 CFR Part 50)." 6
B. COMPARISON OF CLINCH RIVER BREEDER REACTOR DESIGN BASIS ACCIDENTS WITH THOSE FOR LIGHT WATER REACTOR 5 The comparison of Clinch River Breeder Reactor (CRBR) design basis accidents (DBAs) with those for light water reactors (LWRs) is instructive because it illustrates the unique features of the CRBR and because it provides a frame of reference that is probably more familiar to those previously involved in safety and licensing of LWRs. Moreover, such a comparison will also provide insight into the efficacy of the deterministic approach for accident delineation for the CRBR. The comparison is based on Chapter 15 of NUREG-0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Piants, LWP. Edition, July 1981 and Chapter 15 of the Preliminary Safety Analysis Report (PSAR) for the Clinch River Breeder Reactor Project. For each event given in NUREG-0800 Chapter 15, a corresponding event or events given in Chapter 15 of the CRBR PSAR is described in this section with corresponding paragraph numbers to allow direct reference. If there is no direct correspondence, an explanation is presented. Table 1 presents a summary of the comparisons. 7
TABLE 1. SufiMARY Of C0f4PARISON Of CRBR PSAR CllAPTER 15 DESIGN BASIS ACCIDENT EVLHIS Willl ill0SE f'JR LWRs AS GIVEN IN NUREG-0800, STAllDARD REVIEW PL AN Corresponding or Analogous Event in CRBR flVREG-0800 SRP (From PSAR Chapter 15) 15.1 Increase in lleat Removal by the Secondary System (Results in Reactivity insertion) 15.1.1 Decrease in feedwater Teinperature These events are not specifically addressed in PSAR Chapter 15. They are protected against by the steam-15.1.2 Increase in feedwater flow feedwater flow mismatched system and/or the steam drie level detection system. Typically, 15.1.3 Increase in Steam flow and inadvertent perturbations in the steam generating system of the Opening of a Steam Generator Relief CRBR are not reflected in the primary system until Valve long after the reactor has been shut down. 15.1.4 Increase in Steam Flow an inadvertent Opening of a Steam Generator Safety w Valve 15.1.5 Steam System Piping Failures inside There is no steam system piping in the containment in and Outside of Containment (PWR) CRBR. Analogous events for CRBR are: 15.3.3.1 Steam or feed Line Break and 15.7.3.5 f t.el Rod Leakage 15.1.6 Appendix A--Radiological Consequences ) Combined with IHX and Steam Generator Leakage. of Main Steam Line f ailures Outside Containment of a PWR j 15.2 Decrease in lleat Removal by the Secondary System (Results in increased System Pressure) 15.2.1 Loss of External Load The analogous events in CRBR are: 15.3.1.5 Turbine 15.2.2 Turbine Trip ) 15.2.3 Loss of Condenser Vacuum s
l TABLE 1. (continued) l Corresponding or Analogous Event in CRBR fiUREG-0800 SRP (from PSAR Chapter 15) 15.2.4 Closure of Main Steam Isolation Valve The analogous events in CRBR are: 15.3.1.5 Turbine (BWR) Trip and 15.3.2.4 Failure of the Steam Bypass System I k 15.2.5 Steam Pressure Regulator failure l i (Closed) s 15.2.6 Loss of Non-emergency AC Power to the The analngous events in CRBR are: 15.3.1.1 Loss of i Station Auxiliaries Off-Site Electrical Power and 15.7.2.3 Generator Breaker Failure to Open at Turbine Trip { 15.2.7 Loss of Normal feedwater flow The ar.alogous events in CRBR are: 15.3.1.6 Loss of l Normal Feedwater, 15.3.1.4 Inadvertent Closure of One Evaporator or Superheater Module Isolation Valve and 15.3.1.7 Inadvertent Actuation of the Sodium-Water u) Reaction Pressure Relief System. 15.2.8 feedwater System Pipe Breaks inside The analogous event in CRBR is 15.3.3.1 Steam or feed l' and Outside Containment (PWR) Line Pipe Break. There are no feedwater lines in ' containment in CRBR. 15.3 Decrease In Reactor Coolant System Flow Rate (Results in Power / Flow Mismatch-High fuel Tanperature/ Damage) 15.3.1 For BWRs, Partial and Conplete Pump The analogous events in CRBR are: 15.3.1.2 Spurious Trips and Malfunctions of the Recircu- Primary Pump Trip, 15.3.1.3 Spurious intermediate lation Flow Controller to Cause Pump Trip,15.3.2.1 Single Primary Pump Seizure, Decreasing Flow, and 15.3.2.2 Intermediate Pump Seizure.
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t TABLE I. (continued) 1 i j- Corresponding or Analogous Event in CDBR NilREG-0800 SRP (From PSAR Chapter 15) 15.3.2 For PWRs, partial and complete reactor' The analogous events in CRBR are: 15.3.1.2 Spurious coolant pump trips. Primary Pump Trip, 15.3.1.3 Spurious Intermediate Pump Trip, 15.3.2.1 Single Primary Pump Seizure, and 15.3.3 Reactor Coolant Pump Rotor Seizure 15.3.2.2 Intermediate Pump Seizure 15.3.4 Reactor Coolant Pump Shaft Break 15.4 Reactivity and Power Distribution Anomalies (Challen Limits) ges Safe Acceptable Fuel Design s 15.4.1 Uncontrolled Control Rod Assembly The analogous events in CRBR are: 15.2.1.1 Control Withdrawal From a Subcritical or Low Assembly Withdrawal at Startup and 15.2.3.4 Control Power Startup Condition Assembly Withdrawal at Startup-Maximum Mechanical g Speed. 15.4.2 Uncontrolled Control Rod Assembly The analogous events in CRBR are: 15.2.1.2 Control Withdrawal at Power Assembly Withdrawal at Power and 15.2.3.5 Control Assembly Withdrawal at Power-Maximum Mechanical Speed. 15.4.3 Control Rod Misoperation (System Mal- The analogous event in CRBR is 15.2.2.3 Maloperation function or Operator Error) of Reactor Plant Controllers. i 15.4.4 Startup of an inactive Loop or The analogous event in CRBR is: 15.2.3.1 Cold Sodium Recirculation Loop at an incorrect insertion. Temperature 15.4.5 Flow Controller Malfunction Causing The analogous event in CRRR is: 15.2.3.1 Cold Sodium an increase in HWR Core Flow Rate Insertion. i
~ _ _ - _ - _ . __ __ . _ _ _ _ _ _ _ _ .
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l I 1 i TABl.E 1. (continued) - Corresponding or Analogous Event in CRBR HUREG-0800 SRP (from PSAR Chapter 15) 15.4.6 Chemical end Volume Control System There is no analogous event for CRBR since the Malfunctian That Results in a Decrease coolant is not horated. In Boron Concentration in the Reactor Coolant (PWR) 15.4.7 Inadvertent Loading and Operation of a Section 15.4 Local Failure Events r. resents extensive fuel Assembly in an Improper Position, analyses of fuel, control and blanket pin failure ch'racteristics, a behavior and consequences. The analytic approach is used to compensate for the limited in-core instrumentation in CRBR, t 15.4.8 Spectrum of Rod Ejection Accidents Hod ejection accidents for are not possibis in (PWR) CRBR because it is a low pressure system an analogous event is 15.2.2.1 Loss of Ilydraulic floid-4
~
down. Other reactivity insertion events analyzed for CRBR are: 15.2.1.3 Seismic Reactivity insertion
. Operational Basis Earthgrade,15.2.1.4 Snall Reactivity Insertions, 15.2.2.2 Sudden Core Radial Movement, 15.2.3.2 Gas Bubble Pas 7. age Through fuel, Radial Blanket and Control Assens t ies, and 15.2.3.3 1 Seismic Reactivity Insertion Safe Shutdown
! Earthquake. 15.4.8 Appendix A--Radiological Consequences for the reactivity insertion events above there are-of a Control Rod Ejection Accident no radiological consequences since none of the events (PWR) challenge the safe acceptable fuel design its 4ts and fuel pin failures are not predicted. 15.4.9 Spectrum of Rod Drop Accidents (BWR) As for 15.4.8 and 15.4.8 Appendix A (above). l 1 i
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1ABLE 1. (continued) Corresponding or Analogous Etent in CRBR HUREG-0800 SRP (from PSAR Chapter 15) 15.4.9 Appendix A Radiological Consequences As for 15.4.8 and 15.4.ft Appendix A (above). of Control Rod Drop Accident (BWR) 15.5 Increase in Reactor Coolant Inventory . (Challenges to System Integrity or Safe Acceptable Fuel Design Limits) 15.5.1 BWRs--Inadvertent operation of the There are no analogous events for the CRBR because high pressure core spray, high pres- high pressure emergency core cooling systems are not sure coolant injection, or reactor required in CRBR since coolant leaks would not caute core isolation cooling system. depressurization and vapor formation as in LWRs. In v 15.7.I.3 IHX Leak the applicants state that leakage 15.5.2 PWRs--Inadvertent operation of high would be into the primary system and would be pressure emegency core cooling system detected by level indications in the reactor (high pressure injection system) or a overflow tank. malfunction of the chemical and volume control system. s 15.6 Decrease in Reactor Coolant Inventory (Challenges to SAfDL) 15.6.1 Inadvertent Opening of a PWR Prcs- There are no analogous events for CRBR since it is a surizer Rel Mf Valve or a BWR low pressure system. Pressure Relief Valve 15.6.2 Radiological Consequences of the There is no analogous event for CRBR since there are Failure of Small Lines Carrying no primary coolant lines outside containment. Primary Coolant Outside Containment. S t
, e .
9 i ( TABLE 1. (continued) ? Corresponding or Analogous Event in CRBR NilREG-0800 SRP (from PSAR Chapter 15) 15.6.3 Radiological Consequences of Steam The steam generators in CRBR are separated from the Generator Tube failure (PWR) primary system by the' intermediate heat transport system. The applicants analyzed a somewhat analogous event, 15.7.3.5 fuel Rod Leakage Combined with IllX and Steam Generator Leakage, and conclude that the radiological consequences would be insignificant.
! 15.6.4 Radiological Consequences of Main There is no analogous event for the CRBR since, in Steam 1.ine failure Outside Containment CRBR, the steam generation system is separated from (BWR) the primary system by the intermediate heat transport system whereas in the BWR the main itcaid line contains primary system steam.
15.6.5 Loss-of-Coolant Accidents Resulting In CRBR the manifestations of a primary system leak
, from Spectrum of Postulated Piping are different than for an LWR since there is no w Breaks Within the Reactor Coolant coolant depressurization and vapor formation in the Pressure Boundary CRBR. There is no challenge to the SAFDL. A some-what analouous event, 15.3.3.4 Primary lleat Transport l System Pipe Leak was analyzed and based on leak-1 before-break and leak detection assumptions it was j conclud
- 90 significant core temperature transient would occur. ;
15.6.5 Appendix A--Radiological Consequences in 15.6.1.4 Primary lleat Transport System Piping of a Design Basis Loss-of-Coolant Leaks, the applicants conclude from their analyses Accident including Containment Leakage that the radiological consequences of design basis l Contribution. leaks are small fractions of 10 CFR 100 guidelines. l 1 i j i 4
_ . - - _ . ._. - _ _ _ = _ - - .- . . - . _ ~ _ , _ _ _ - _ - _ . . . _ - . _ _
)
TABLE 1. (continued) i Corresponding or Analogous Event in CRBR fluREG-0800 SRP (from PSAR Chapter 15) 15.6.5 Appendix B--Radiological Consequences There is no directly analogous event for the CRBR of a Design Basis Lc.5-of-Coolant since all the primary piping and associated ! i Accident: Leakage from Engineered engineered safety features (reactor cavity and cell Safety Components Outside Containment liner) are within containment. 15.6.5 Appendix C--Radiological Consequences of a Design Basis Loss-of-Coolant Accident: Post-LOCA Purge Contri-1- bution--DELETED. i 15.6.5 Appendix D--Radiological Consequences There is no directly analogous event for the CRBR. ! of a Design Basis Loss-of-Coolant CRBR systems are more akin to PWR systeins. Accident: Leakage froin Main Steam isolation Valve Leakage Control System (BWR) U 15.7 Radioactive Release from a Subsystem Component 15.7.1 Waste Gas System Failure--DELETED. 15.7.2 Radioactive Liquid Waste System Leak or failure (Release to Atmosphere)-- DELETED. 15.7.3 Postulated Radioactiire Releases Due To Analogous events in CRBR are: 15.6.1.2 Failure of Liquid-Containing Tank Failures. the Ex-Vessel Storage Tank Sodium Cooling System Dur-ing Operation, 15.6.1.3 Failure of an Ex-Containment Primary Storage Tank, and 15.7.2.5 Liquid Radwaste ! System Failure (Leak or Rupture). i
- s
- TABLE 1. (continued)
Corresponding or Analogous Event in CRBR fiUREG-0800 SRP (from pSAR Chapter 15) 15.7.4 Radiological Consequences of Fuel Analogous events for CRBR are: 15.5.2.1 Fuel I llandling Acc idents. Assembly Dropped Within Reactor Vessel During Refueling, 15.5.2.2 Damage of fuel Assembly Due To Attempt To Insert A fuel Assembly Into An Occupied position, 15.5.2.3 Single fuel Asembly Cladding failure and Subsequent Fission-Gas Release During
)
Refueling,15.5.2.4 Cover-Gas Release During Refueling,15.5.3.1 Collision of EVlM with Control Rod Drive Mechanisms, and 15.5.2.5 The lleaviest < Crane Load Impacts the Reactor Closure llead. 15.7.5 Spent fuel Cask Drop Accidents The analogous event in CRBR is 15.7.3.2 Spent fuel Shipping Cask Drop From Maximum possible lleight. 15.8 Anticipated Transients Without Scram G 15.8.1 Anticipated Transients Without Scram The analogous events in CRBR are called hypothetical core disruptive accidents. Because they are not con-sidered design basis accidents they are not addressed in PSAR Chapter 15. 4 i 4
t 15.1 Increases in Heat Removal By The Secondary System In this family of DBAs the principal concerns for the LWRs are "a decrease in moderator temperature which increases core reactivity and can i lead to a power level increase and a decrease in shutdown margin." "The power level increase will lead to a reactor trip."
"Any unplanned power level increase may result in fuel damage or excessive reactor system ~
i pressure." There are several important characteristics of the CRBR which minimize j the effects of changes in heat removal' rate in the steam side of the j plant: the primary system pressure is insensitive to the coolant i temperature because it is controlled by the reactor cover gas pressure control system; the sodium outlet temperature is normally several hundred Fahrenheit degrees below saturation; the temperature transient transport
- time from the steam side to the primary side (through the intermediate heat transport system) is relatively long with respect to reactor scram time, j the reactivity effect of coclant temperature changes is smaller in CRBR than for corresponding coolant temperature change in an LWR, and, because
; the margin between normal operation and fuel damage limits is so large, in j many instances the effects of temperature transients on plant component lifetimes are more limiting than those of the fuel, i
15.1.1 Decrease in Feedwater Temperature ' 4 15.l.2 Increase in Feedwater Flow Although not specifically addressed in Chapter 15 of the CRBR PSAR these transients would not challenge reactor safety from the standpoint of reactivity increase, primary system pressure or fuel damage limits. Feedwater temperature is monitored by three resistance temperature . detectors in the steam drum inlet line. The RTO signals are used to j provide temperature compensation for the feedwater flow signal. They are , also displayed in the control room. Feedwater mass flow is sensed by three _ j differential pressure elements across one venturi in the inlet line to each i j 16
steam drum. These temperature-corrected feedwater flow signals are supplied to the reactor shutdown system logic. The reactor shutdown system
; provides buffered signals to the plant control system and the plant data '
handling and display system. Steam drum level is sensed by three differential pressure elements measuring the differential pressure between. a reference column and the water head in the steam drum. This measurement I is density compensated. The signal is supplied to the reactor shutdown system logic. Buffered signals are supplied to the plant control system and the plant data handling and display system. The steam-feedwater flow mismatch subsystem initiates reactor trip to prevent continued operation with large imbalances between the steam and feedwater flow for each heat j transport system loop. These subsystems protect the steam generators and drums against unacceptable thermal transients. Each subsystem compares the steam and feedwater flow in two individual comparators. If the difference between the two values exceeds the setpoint in either of the camparators, a l trip is initiated. Increasing steam flow and decreasing feedwater flow fault events are sensed by the first comparator. The second comparator senses decreasing steam flow and increasing feedwater flow fault events. j A decrease in feedwater temperature would likely manifest itself by a change in steam drum level and/or a change in the feedwater mass flow rate (since the flowmeter signals are temperature compensated). Both process I variables supply signals to the reactor shutdown system. An increase in feedwater flow rate would be detected by the steam-feedwater ficw mismatch system and/or the steam drum . level sensors. f Both process variables supply signals to the reactor shutdown system, i As discussed earlier, the reactor would be scrammed long before the temperature transient is reflected back into the primary heat transport system and no challenge is presented to the primary system or the fuel. I
. 15.1.3 Increase in Steam Flow and Inadvertent Opening of a Steam Generator Relief Valve b l i
l 4 17
--nv -r-n-,-e.w--,-- .,,-r----.,..--se ---,-~,-,---,,-v--,,--,---=m-nn~ -----+---w-.,v- . , - - , -w.---,.----,-----m- r-m-- w, r e ve --m e n e , -, ~,
15.1.4 Increase in Steam Flow and Inadvertent Opening of a Steam Generator Safety Valve In these events reactor trip occurs due to steam-feedwater flow mismatch or low steam drum level. Again, the reactor is shutdown long before the thermal transient manifests itself in the primary system hence, there are no challenges to the primary system or fuel damage limits. - 15.1.5 Steam System Pioino Failures Inside and Outside of Containment (PWR) 15.1.5 Appendix A Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR In the CRBR there is no steam system piping in containment. Analogous events for CRBR are 15.3.3.1 Steam or Feed Line Pipe Break and 15.7.3.5. Fuel Rod Leakage Combined with IHX and Steam Generator Leakage. Again, the reactor is shutdown long before the thermal transient is manifested in the primary system. The radiological consequences of a steam system piping break are very small (insignificant) because pressure differentials between the steam system and the intermediate heat transport system and the primary heat transport ~ system are such that leakage would be into the primary system. 15.2 Decrease in Heat Removal by the Secondary System In this family of DBAs the principal concerns for the LWRs are an increase in primary coolant temperature and pressure which present potential challenges to primary system integrity and fuel damage limits. , As previously discussed, in the CRBR these events do not challenge the primary system integrity or fuel damage limits because the reactor is . scrammed long before the temperature transient is manifested in the primary system. 18
15.2.1 Loss of External Load 15.2.2 Turbine Trio 15.2.3 Loss of Condenser Vacuum The analogous event in CRBR is 15.3.1.5 Turbine Trip. A turbine trip intitiates opening of the steam bypass valves (80% steam bypass capability) and a runback in reactor power (3% per minute). Two hundred seconds after the turbine trip occurs, reactor power has decreased to 90% and is continuing to decrease at 3% per minute. Reactor shutdown is completed after this time. In the event the Steam Bypass System fails, 15.3.2.1.4 the reactor is scrammed by a steam-feedwater flow ratio trip and when the available normal feedwater supply is exhausted (>20 minutes), the Steam Generator Auxiliary Feat Removal System (SGAHRS) is activated bj a low level drum trip and feedwater is provided by the auxiliary feedwater pumps. The resulting core temperatures are very similar to those for a normal trip from full power. Thus, there are no challenges to primary system integrity or fuel damage limits. Long term core cooling is accomplished with the Steam Generator Auxiliary Heat Removal System. 15.2.4 Closure of Main Steam Isolation Valve (BWR) 15.2.5 Steam Pressure Regulator Failure (Closed) Analogous events in CRBR are bounded by 15.3.1.5 Turbine Trip and 15.3.2.4 Failure of the Steam Bypass System (discussed above). 15.2.6 Loss of Nonemergency AC Power to the Station Auxiliaries The principal concerns for the LWRs are challenges to the primarv system integrity and the safe acceptable fuel design limits. The analogous
, events in CRBR are 15.3.1.1 Loss of Off-Site Electrical Power and 15.7.2.3.
Generator Breaker Failure to Open at Turbine Trip. As stated in the PSAR, the applicants conclude, "The loss of off-site electrical power results in 19
l' : a simultaneous loss of sodium pump power and the consequent reduction in core flow. The primary shutdown system limits the clad (sic) midwall hot spot temperature to 1410 F. In the unlikely event that the primary shutdown system does not operate, the secondary shutdown system limits the hot spot midwall clad (sic) temperature to 1630 F. Tnis is an acceptable ' i result because analysis of the transient has shown that the cladding damage (cumulative damage function) does not exceed the limit for an emergency , I event." 15.2.7 Loss of Normal Feedwater l In this event the principal concerns for the LWRs are increases in 1 coolant pressure and temperature which could challenge primary system integrity and the safe acceptable fuel design limits. This event for CRBR is analyzed in 15.3.1.4 Inadvertent Closure of One Evaporator or Superheater Module Isolation Valve,15.3.1.6 Loss of Normal Feedwater, and 15.3.1.7 Inadvertent Actuation and the Sodium-Water Reaction Pressure Relief System. The applicants conclude from analysis of these events that " core temperature following the loss of normal feedwater supply is equivalent to the temperature following a normal plant trip, and large j margins are available in the active systems required for auxiliary ! cooling. The reactor and heat transport systems are designed to accommodate this event." 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment (PWR) The principal concerns for the PWR in this event are primary system integrity, safe acceptable fuel design limits, continued core cooling, and radiological consecuences. I
- I i
For the CRBR it should be noted that there are no steam or feedwater lines in the containment. This event is analyzed in 15.3.3.1 Steam or Feed . Line Pipe Break. As stated previously, the temperature transient is not ; manifested in the primary system until after the reactor has been l 20 i t .. _ _ _ _ _ _ . . . - _ _ _ _ _ - - . _ _ _ _ _ _ . _ ._ ____,-_.--_. _ _ _ . . _ . _ , - - . . ~ . - - - . . . _ ,
scrammed. The applicants analyzec a number of postulated break locations. Based on the worst break location, saturated steamline between the steam drum and the superheater, they conclude that no challenges are presented to primary system integrity, fuel design limits, continued core cooling and 10 CFR 100 limits. 15.3 Decrease in Reactor Coolant System Flow Rate Tne principal concerns for LWRs for these DBAs are primary system integrity, reactivity control and safe acceptable fuel design limits. Specific events are: 15.3.1 For BWRs, Partial and Complete Pump Trios and Malfunction of the Recirculation Flow Controller to Cause Decreasing Flow 15.3.2 For PWRs Partial and Complete Reactor Coolant Pump Trios 15.3.3 Reactor Coolant Puno Rotor Seizure 15.3.4 Reactor Coolant Pump Shaf t Break The analogous events for CRBP are: 15.3.1.2 Spurious Primary Pumo Trio 15.3.1.3 Spurious Intermediate Pump Trip 15.3.2.1 Sinole Primary Pump Seizure 15.3.2.2 Single Intermediate Pump Seizure 1 Core temperatures are as for a normal plant trip. There are no
. challenges to primary system integrity, reactivity control or fuel design limits.
21
J 15.4 Reactivity and Power Distribution Anomalies 1 The principal concerns for LWRs for these events are challenges to the safe acceptable fuel design limits. 2 15.4.1 Uncontrolled Control Rod Assembly Withdrawal From a Subcritical or Low Power Startup Condition , The analogous events in CRBR are 15.2.1.1 Control Assembly Withdrawal
- at Startup and 15.2.3.4 Control Assembly Withdrawl at Startup-Maximum ,
Mechanical Speed. No challenges are presented to the fuel design limits. 15.4.2 Uncontrolled Control Rod Assembly Withdrawl at Power The analogous events in CRBR are 15.2.1.2 Control Assembly Withdrawal at Startup and 15.2.3.5 Control Assembly Withdrawal at Power-Maximum I Mechanical Speed. Although fuel pin cladding temperatures rise above normal operating temperatures the applicants conclude from their analyses that the safe acceptable fuel design limits are not challenged. I 15.4.3 Control Rod Misoperation (System Malfunction or Operator Error) i The analogous event in CRBR is 15.2.2.3 Maloperation of Reactor Plant
- Controllers. The reactivity ramp is terminated by the plant protection
! system within the operational incident limits.
15.4.4 Startup of an Inactive Loop or Recircu,lation Loop at an Incorrect Temperature The analogous event for the CRBR is 15.2.3.1 Cold Sodium Insertion.
- The applicants have classified this event as an extremely unlikely event whereas the Standard Review Plan considers such an event to be of " moderate
! frequency" for LWRs. . I l l 22 r
15.4.5 Flow Controller Malfunction Causing an increase in BWR Core Flow Rate The principal concern with this event in a BWR is the resultcr.t positive reactivity addition effect. The corresponding, although extremely unlikely event for the CRBR is 15.2.3.1 Cold Sodium Insertion. 15.4.6 Chemical and volume cceprol System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant (PWR) There is no analogous event for the CRBR since the coolant in CRBR is not borated for reactivity control. 15.4.7 Inadvertent Loadino and Operation of a Fuel Assembly in an Improper Position The principal concerns for LWRs for this event are the detectability of a misloaded fuel assembly after fueling operations or, failing that, offsite radiological consequences. Section 15.4, Local Failure Events, in the CRBR PSAR presents extensive analyses of fuel, control and blanket pin failure characteristics, behavior and consequences. The principal impetus for these analyses is the fact that there is only limited in-core instrumentation in CRBR so immediate detection of misloaded or overenriched pins is not possible. Major categories of events analyzed are: 15.4.1.1 Stochastic Core Fuel Pin Failure 15.4.1.2 Overenriched Fuel Rod Failure i 15.4.1.3 Flow Blockage in a Core Assembly 15.4.2.1 Stochastic Absorber Pin Failures 23 I
15.4.2.2 Overoower Control Rod Assembly 15.4.2.3 Flow Blockage of a Control Rod Assembly 15.4.3.1 Stochastic Radial Blanket Pin Failure 15.4.3.2 Overpower Radial Blanket 15.4.3.3 Flow Blockage of a Radial Blanket Assembly 15.4.8 Spectrum of Rod Ejection Accidents (PWR) The principal concerns in LWRs for this event are integrity of the primary system, safe acceptable fuel design limits, and radiological consequences. Because the primary coolant side of CRBR is not a high pressure system, control rod ejection accidents per se are not possible. A somewhat analogous event is 15.2.2.1 Loss of Hydraulic Holddown. Other reactivity insertion events analyzed for the CRSR include 15.2.1.3 Seismic Reactivity Insertion-Operational Basis Earthquake, 15.2.1.4 Small Reactivity Insertions,15.2.2.2 Sudden Core Radial Movement,15.2.3.2 Gas Bubble Passage Through Fuel, Radial Blanket and Control Assemblies, and 15.2.3.3 Seismic Reactivity Insertion-Safe Shutdown Earthquake. 15.4.8 Appendix A Radiological Conseauences of a Control Rod Ejection Accident (PWR) For the design basis reactivity insertion accidents in CRSR just discussed there are no radiological consequences since none of the events challenge the safe acceptable fuel design limits, and fuel pin failures are not predicted. 15.4.9 Spectrum of Rod Drop Accidents (BWR) 24
15.4.9 Appendix A, Radiological Consecuences of Control Rod Droo Accident (BWR) Analogous events for the CRBR were discussed above. 15.5 __ Increase in Reactor Coolant Inventory The principal concerns for LWRs from these events are challenges to the primary system integrity and the safe acceptable fuel design limits. 15.5.1 BWRs-Inadvertent operation of the high pressure core spray, high pressure coolant injection, or reactor core isolation cooling system 15.5.2 PWRs-Inadvertent operation of high pressure emergency core cooling system (high pressure injection system) or a malfunction of the chemical and volume control system There are no analogous events for the CRBR because there are no high pressure emergency core cooling systems in the CRBR. These systems are not required in the CRBR because coolant leaks do not cause depressurization of the system and vapor formation. 15.6 Decrease In Reactor Coolant Inventory The principal concerns in LWRs for these events are challer.ges to the safe acceptable fuel design limits and primary system integrity. The CRBR primary coolant system is a low pressure system and the coolant under normal conditions is several hundred Fahrenheit degrees below its saturation temperature. Hence, postulated leaks in the system do not result in depressurization or coolant boiling as ir. LWRs. Moreover, the CRBR primary heat transport system is designed to preclude uncovering the i core if there is a break in the hot leg piping. In addition, the applicants contend that the piping will leak before it breaks, thus providing ample time for reactor shutdown and isolation of the leak. 25 r - - ~ ~ e o m e v <-- -
. - ~_
+ 1 15.6.1 Inadvertent Opening of a PWR Pressurizer Relief Valve or a BWR : Pressure Relief Valve 4 There is no analogous event for the CRBR since the primary system is a low pressure system (sl atmosphere at the top of the sodium in the
~
i reactor vessel). 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment j There are no primary coolant lines outside containment in the CRBR. 15.6.3 Radiological Consequences of Steam Generator Tube Failure (PWR) i The steam generators in the CRBR are separated frcm the primary system by the intermediate heat transport system. The applicants.have analyzed a j somewhat analogous event, 15.7.3.5. Fuel Rod Leakage Combined with IHX and
- Steam Generator Leakage and conclude that the radiological consequences would be insignificant.
15.6.4 Radiological Consequences of Main Steam Line Failure Outside Containment (BWR) There is no analogous event for the CRBR since in CRBR the steam generation system is separated from the primary system by the intermediate heat transport system whereas in the BWR the main steam line contains i primary system coolant (steam). 15.6.5 Loss-of-Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within The Reactor Coolant Pressure Boundary i { Although the applicants contend that an instantaneous pipe break in l CRBR is incredible, in 15.3.3.4 Primary Heat Transport System Pipe Break, . they present the results of analyses to determine the effect on core temperatures of pipe leaks. They conclude that a leak of 75,000 gallons 26
per minute would be necessary for core sodium temperatures to approach saturation. They estimate that leak detection capability will be better ! than three gallons per minute and that a 30 gallon per minute leak would not result in a measurable core temperature transient, would permit a normal' reactor shutdown and provide several days for further reduction of the leakage. In 15.3.3.2 Loss of Normal Shutdown Cooling System, long term core cooling is accomplished using the steam generator auxiliary heat removal system. TL? consequences of leaks in the intermediate heat transport system are evaluated in 15.3.3.5 Intermediate Heat Transport System Pipe Leak. 15.6.5 Appendix A. Radiological Consecuences of a Design Basis Loss-of-Coolant Accident Including Containment Leakage Contribution In 15.6.1.4 Primary Heat Transport System Piping Leaks, the applicants analyze the radiological consequences of sodium leaks from the primary system (and the resultant aerosols from sodium combustion) and conclude that they are small fracticns of 10CFR100 guideline values. 15.6.5 Appendix B, Radiological Consecuences of a Design Basis Loss-of-Coolant Accident Leakage From Engineered Safety Components Outside Containment There is no directly analogous event 'or the CRBR since all the primary piping and associated engineered sifety features (reactor cavity and cell liners) are within containment. In 15.6.1.5 Intermediate Heat Transport System Pipe Leak, the applicants analyze the radiological consequences of a leak in the IHTS with an assumed leak in the IHX. They conclude that the radiological consequences are below 10CFR100 guideline limits. l 4 27
15.6.5 Aopendix C, Radiological Consequences of a Design Basis Loss-of-Coolant Accident, Leakace From Main Steam Isolation Valve Leakace Control System (BWR) There is no analogous event for the CRBR (see the discussion for the last two previous events). 15.7 Radioactive Release From a Subsystem or Component 15.7.3 Postulated Radioactive Releases Oue To Liquid-Containing Tank Failures a The principal thrust of this item for LWRs is the evaluation of the radiological consequences of failure of tanks outside containment which contain radioactive liquids. For CRBR the analogous events are: 15.6.1.2 Failure of The Ex-Vessel Storage Tat,4 Sodium Cooling System During Operation,15.6.1.3 Failure of an Ex-Containment primary Sodium Storage Tank, and 15.7.2.5 Liquid Radwaste System Failure (Leak or Rupture). 15.7.4 Radiological Consequences of Fuel Handling Accidents The analogous events for CRBR are 15.5.2.1 Fuel Assembly Dropped Within Reactor Vessel During Refueling, 15.5.2.2 Damage of Fuel Assembly Due to Attempt to Insert a Fuel Assembly Into An Occupied Position, 15.5.2.3 Single Fuel Assembly Cladding Failure and Subsequent Fission-Gas Release During Refueling,15.5.2.4 Cover Gas Release During Refueling, 15.5.3.1 Collision of EVTM With Control Rod Drive Mechanisms, and 15.5.2.5 The Heaviest Crane Load Impacts The Reactor 'losure Head. 15.7.5 Spent Fuel Cask Drop Accidents The analogous event in CRBR is 15.7.3.2 Spent Fuel Shipping Cask Drop From Maximum Possible Height. ,
- Note: Items 15.7.1 and 15.7.2 have been deleted from the SRP (NUREG-0800).
28
.- - _ - _ ~ - - - -. ._ - _ _ _ . __ . -_ .- - -
4 15.8 Anticipated Transients Without Scram Anticipated transients without scram (ATWS) are those low probability h events in which an anticipated transient occurs and is not followed by an j automatic reactor shutdown (scram) when required. Concern for ATWS arises j . principally from consideration of common mode failures in the reactor protection system. Reduction of the probability for common mode failure is i obtained by employing redundancy and diversity in the reactor protection system. In the case of light water reactors this generally had to be accomplished on a retrofit basis. In the case of the CRBR, reactor protection system diversity and redundancy received a great deal of attention from the outset. This is reflected in the design of the ! protection systems which include two separate and diverse sets of absorber rods, each capable of shutting down the reactor. Because particular attention has been paid in the CRBR design to prevent ATWS events, the probability of an ATWS event is deemed to be extremely small. The consequences of an ATWS event, which in fast reactor parlance, is called an ! hypothetical core disruptive accident, are analyzed in great detail. These l analyses are incorporated by reference in the PSAR (Section 1.6). l l C. DESIGN BASIS ACCIDENTS UNIQUE TO CRBR 1 From the foregoing one may conclude that there is generally good correlation between DBAs for LWRs as delineated in the Standard Review Plan and in the CRBR PSAR. However, a number of DBAs are unique to the CRBR, i which are quoted in this section from the CRBR PSAR. The reader should bear in mind that in the following DBA discussions the conclusions
- regarding the consequences of a particular DBA are those of the applicants.
l 15.3.2.3 Small Water-to-Sodium Leaks In Steam Generator Tubes. A water-to-sodium leak detection system is provided to alert the operator to
, leaks as small as a 2 x 10-5 lb water /sec. For intermediate and large
! leaks the reactor will be shut down normally folicwed by a controlled l - cooldown and depressurization of the affected steam generator. l 29
i 15.3.3.3 Large Sodium-Water Reaction. A large leak in a steam l generator tube will result in injection of high pressure steam and/or water into the IHTS sodium. The resulting sodium-water reaction (SWR) will J generate higher than normal pressures and temperatures in the IHTS. As discussed in Section 15.3.2.3, Steam Generator Tube Leak, the probability of a leak in a tube in the steam generators is expected to * - 7 11 as a result of careful design supported by development and testing of the , steam generators. dowever, a leak detection system, described in Section 7.5.5, has been provided to allow operator action to limit the consequences of a leak in a steam generator tube. The leak detection system will alert the operator to the existence of a leak rate as low as 2 x 10 -5 lb water /sec. For initial leak sizes which can be realistically expected (up to about 10- lb water /sec) there will be sufficient time j for operator action to limit damage to the steam generator and to prevent a significant increase of the leak rate. Should a leak occur of such , magnitude that operator action as described above is not effective, the Sodium-Water Reaction Pressure Relief Subsystem (SWRPRS) will provide sodium side pressure relief by oneration of the rupture discs in the IHTS so that integrity of the IHTS piping and components, e.g., pump and the Intermediate Heat Exchanger (IHX) will be maintained. No operator action is required for the SWRFRS to perform its design function. A description i of the SWRPRS is given in Section 5.5. 15.6.1.1 Primary Sodium In-Containment Storage Tank Failure During Maintenance. This event, which is discussed in detail in PSAR Section 6.2 i Containment Systems, is the design basis accident for the containment systems. It involves rupture of the crimary sodium storage tank in a deinerted cell and containment and the resultant sodium fire products. 15.7.1.1 Loss of D.C. System. The plant design includes three incependent battery supported Class lE 0.C. systems which are described in , Section 8.3.2.1. The loss of one D.C. system will not prevent the operation of Class lE D.C. loads as these systems will be designed with . sufficient physical separation, electrical isolation, and redundancy to l prevent occurrence of common failure modes. l 30
It should be noted that the loss of one D.C. system will not result in a loss of the associated vital bus as this bus will be automatically transferred to the Class lE A.C. Distribution System. Since the transfer will be accomplished with static transfer switching circuitry synchronized with the aforementioned system voltage, the performance of loads fed from the vital bus will not be degraded by the transfer. 15.7.1.2 Loss of Instrumentation or Valve Air. The systems supplying compressed air to safety-related valves or instruments will be designed such that a single credible failure will not cause interruption of the air i supply. The instrument air system is designed to supply clean, dry, and oil-free air for plant instrumentation and controls. The air receiver tanks are designed to the ASME Boiler and Pressure Vessel Code, i Section VIII, Division 1. Piping is designed to ANSI B31.1.0. Piping which penetrates the reactor containment walls, and the containment isolation valves are designed to ASME Section III, (Sections 3.9.2 and 6.2.4). Intercoolers and aftercoolers 3re designed to Tubular Exchangers Manufacturers Association (TEMA) Class R." i 15.7.1.4 Off-Normal Cover Gas Pressure in The Reactor Coolant Boundary. Off-normal cover gas pressures in the reactor coolant boundary will not cause a safety problem. Underpressure would be limited to approximately 1 psi below the normal operating pressure of 10 inches W.G. Overpressure conditions would be limited to 15 psig by relief actions and would take about an hour to achieve. Even if such an overpressure condition were to exist, there will be no deleterious effect on the i integrity of the reactor coolant boundary. 15.7.1.5 Off-Normal Cover Gas Pressure In IHTS. High cover gas pressure (to the maximum pressure of the argon supply) has no deleterious effect on IHTS performance. Lower cover gas pressure (to atmospheric pressure) will have no effect on IHTS heat transport capability. Low pressure could result in the introduction of radioactive material into the IHTS through the IHX cnly after f ailure of the IHX intermediate-primary aP monitor, failure of the cover gas pressure indicator, and leaks in the IHX tubes. This would be detected by a radiation monitor in the IHX I intermediate side outiet. ! 31
._- - - _- _. -=
15.7.1.6 Small NaK Spills in The EVST NaK Systems. The absence of radioactivity, coupled with the design of the NaK system and components, as i well as the available fire protection system, provides assurance that any anticipated leakage from the EVST NaK system will not adversely affect EVST cooling. The applicants conclude that there is no potential radiological release from the plant.
- 15.7.2.1 Inadvertent Release of 011 Through Pump Seal (PHTS) It is
} highly improbable that any quantity of oil could be released through the pump seals in such a manner as to interact with the primary sodium cool:nt. Nevertheless, analyses, based on conservative assumptions, have been carried out to determine the resultant; (1) plugging effects and (2) I reactivity effects associated with this. postulated occurrence. The applicants conclusions from these analyses indicate that; (1) during normal operation with mixing in the total primary inventory the plugging
- temperature was found to be on the order of 440*F, well below the minimum
! operating temperature of approximately 640*F, (2) during Refueling or Hot Standby the plugging temperature was found to oe well below 377*F which is below the 400*F temperature for refueling conditions, and therefore presenting no safety problems, and (3) the potential reactivity effect associated with this event is of such a small nature that the consequences to the core are considered insignificant. l 15.7.2.2 Inadvertent Release of Oil Through The Pumo Seal Into Sodium (IHTS). The release of oil in the PHTS has been discussed in
- Section 15.7.2.1. The release of oil to the IHTS from the pump oil bearing I
requires the failure of multiple barriers designed to prevent such a release. If oil contamination of the IHTS sodium did occur, it could be detected by monitoring the seal oil inventories or from a chemical analysis i of sodium samples. An undetected loss of the entire seal oil supply to the IHTS sodium would have consequences for the IHTS heat transport capability no more severe than those evaluated in Section 15.3.2.2 (Single Intermediate Loop Pump Seizure), Section 15.3.3.5 (Intermediate Loop Pipe l Break), or Section 15.7.2.1 (Inadvertent Release of Oil through Pump Seal (PHTS)). 32
l l l 15.7.2.4 Ruoture In RAPS Cold Box. The Radioactive Argon Processing System cold box contains the cryogenic still which separates krypton and xenon from the reactor argon cover gas stream. It is assumed that the reactor has been operating a long time with 1% failed fuel and that the cryostill has not been unloaded to the noble gas storage vessel for one year. Thus, the cryostill contains a maximum inventory of radioactivity. Assuming a cryostill rupture, the cola oox cell H&V radiation monitor will signal an alarm, close off the cold box influent and effluent lines and open the bypass line. The liquid nitrogen will be valved off on a cell high-pressure signal. It is assumed that the RCB refueling door is open and that the ecleased gas vents out of the H&V exhausts, without retention in the cold box cell. The applicants conclude that the maximum individual dose at the site bourdary in 2 hours is 3.0 rem whole body. 15.7.2.6 Failure (Leak or Ruoture) in The EVST NaK System. There are 3 NaK systems for cooling the EVST sodium, 2 active and one natural-draft standby. Each involves a Na-NaK heat exchanger outside the Ex-Vessel Storage Tank and a loop with a NaK/ air heat exchanger using outside air for a heat sink. The NaK systems operate at about 100 psig and 350 F and are all welded construction. Leaks of NaK in the inerted cells containing EVS components will have smaller impacts than those involving EVS sodium because the NaK systems have smaller liquid volumes at lower temperatures and are non-radic:ctive. Leaks in either the air atmosphere cells or the heat exchangers will result in localized fires handled by catch pans, sealing of the volume and extinguishing by the fire protection system or by nitrogen flooding. The applicants conclude that consequences to the public consist of the release to the air of minor amounts of NaK combustion products. - 15.7.2.7 Leakage from Sodium Cold Traos. In the CRBRP auxiliary systems there are the primary sodium cold traps (cooled by liquid NaK in a jacket welded to the outside of the trap), the EVS sodium cold traps (cooled with nitrogen), and the intermediate sodium cold traps (air cooled). All cold traps collect oxide impurities, tritium and other radioactive materials as precipitates in their crystallizer component. A leak between the NaK and Na in a primary system cold trap will leak NaK 33 I l
into the primary sodium; loss of NaK inventory will signal the failure. A leak to cooling gas or to cell atmosphere will be signaled by leak detectors at the coolant outlet or in the cell; the trap will be isolated and replaced. A leak in any of the 3 types of cold traps will have negligible effect on either reactor safety or public safety. When a cold
~
trap is to be removed, it is isolated from the processing system and the sodium allowed to freeze. The trap is cut out and the ends cc,3 ped and seal-welded. Once removed, all traps remain frozen without supplementary cooling. Consequences of a Design Basis Cold Trap Fire (within the RCB with RCB leakage at 0.032% per aay) as calculated by the applicants indicate ; maximum individual dose for 2 hours at the site boundary of 1.02 mrem to the bone. 15.7.2.8 Rupture in RAPS Noble Gas Storace Vessel Cell. The RAPS noble gas storage vessel normally contains radioactive gas which is unloaded annually from the RAPS cryostill. The gas is mainly argon but also includes krypton and xenon; both stable and radioactive isotopes are present. The gas is bled slowly from the vessel into th Cell Atmosphere Processing System so that its pressure normally decreases over the year. The event analysis assumes that shortly after cryostill unloading, it is again unloaded so that the storage vessel contains two charges and is at maximum pressure when it ruptures. A radiation monitor initiates a signal closing off the cell vent line to CAPS. It is assumed that the cell leaks the gases quickly into the RCB, that the RCB refueling door is open, and that the released gases vent out through the H&V exhausts. The applicants estimate the maximum dose to an individual at the site boundary for 2 hours is 3.0 rem whole body. 15.7.2.9 Rupture in the CAPS Cold Box. The Cell Atmosphere Processing System cold box contains two charcoal delay beds in series, which absorb xenon and krypton from the process stream before it is discharged to H&V. The beds are cryogenically cooled by injecting liquid nitrogen into the influent stream. The event analysis assumes a rupture of , the charcoal delay beds during refueling when the bed inventory of radionuclides is greatest. Normally the cell H&V radiation monitor would initiate an alarm and closing of the H&V vent, the tritium water removal i 1 34
drain, and process gas flow. High cell pressure would trip off the liquid nitrogen flow. The applicants analysis of the event assumes all the_ __ _ __ _ { radioactivity is immediately released from the RSB, leading to an estimated j maximum 2-hour site boundary individual dose of 0.14 rem whole body, i
. 15.7.3.1 Leak in a Core Component Pot. A Core Component Pot is a long stainless steel thimble used to hold a fuel assembly or control assembly in transferring it from the reactor vessel to the EVST or reverse. A CCP contains about 22 gallors of sodium in addition to the core
- component; the plant may have 600 or more CCPs. A CCP could fail due to
- defective manufactura or to accidental damage or by corrosion. The l limiting event analyzed is undetected loss of sodium from the CCP
{ immediately before transporting a 20 KW-decay-heat fuel assembly in the CCP j in the Ex-Vessel Transfer Machine. Normally, the time between a CCP emerging from the reactor sodium surface and submerging be~ neath the EVST sodium surface is 56 minutes. The event analysis assumes a fuel urloading procedure unperturbed by low-load signal from the grapple load cell, or by the monitor signalling high radiation levels in the EVTM. After 10 minutes, the fuel would heat to cladding temperatures of 1500 F and random cladding failure could release some fission gas; the EVTM has a 20-kW capacity gas flow cooling system but the fuel would be insulated by the still gas between it and the CCP. After about 17 minutes, cladding begins'to melt; after 30 minutes it is melting on all rods. After about an
; hour, the fuel assembly duct reaches the melting point locally. Even after 1.5 hours, the CCP (at 1900*F) would maintain its integrity. It is assumed that all the volatile fission products are released to the inside of the EVTM (if they are molten or vapor at 3500 F). The release is assumed to be by diffusion through the EVTM seals (calculated to reach 260 F) to the RSB/RCB and to involve only radionuclides mobile at 260 F. The results are said to be represented by those of Section 15.5.2.3, in which the maximum 2-hour site boundary dose to an individual is estimated at 1.89 rem to the
. thyroid.
6 4 35 1
- ----,,-m-. _ - - - - - , - - , - , - - ,, , ,m ,- ,,-w---.--,-e- -- ,--n- +rermn - + +,m-e v -e q g - - - , - - - - - -- - - +r
15.7.3.3 Maximum Possible Conventional Fires, Flood, Storms or Minimum River Level Fire: The plant fire protection system is designed to provide-adequate fire protection, detection and signalling in areas where a hazard mdy exist; the nearest tree line is kept 100 yards away to minimize forest fire impact. _ Flood: Plant grade is at the 815 foot elevation while maximum flood is estimated at 809.2 feet including combined 1/2 Probable Maximum Flood, OBE with dam failures and wave runup. Storms: Drainage facilities for safety-related structures are designed for a maximum 1-hour rain of 14 inches and an 8-hour depth of 29.5 inches; maximum recorded is 7.75 inches in 24 hours. Roof design is > for 40 inches of snow, compared to a maximum monthly snow f all of 21 inches. Wind design for safety-related structures is for 90 mph at 30 feet above grade, compared to a peak recorded gust of 59 mph at Oak Ridge. Tornado design is for a maximum velocity of 360 mph. Minimum River Level: The minimum river level is taken as 735 feet elevation, and the intake structure is 5.5 feet lower. 15.7.3.4 Failure of Plug Seals and Annuli. A transient causing excessively low cover gas pressure at the reactor head, followed by a return to normal, could displace the liquid in the plug annulus dip seals enough to allow cover gas under the seal blade and up into the inner buffer annulus between the dip seal and the inflatable seals. This could increase the rate of leakage of radioisotopes into .ne head access area; factors against this being significant are 1) there are 2 outer saals with a buffer space between them and these seals have more than adequate pressure , resistance, 2) expected leak rates cause a small fraction of the allowable l concentrations in the head access area, and 3) a radiatien monitor in the , head access area can signal any need for evacuation and isolation of containment. l l 36 i
15.7.3.6 Sodium Interaction with Chilled Water. For an interaction l between sodium and water of the Chilled Water Systems to occur, two pipe failures and a third boundary must fail simultaneously. In recirculating gas fan cooler units and HVAC coolers serving areas containing sodium piping or equipment, water leakage will trigger redundant moisture or
. leakage detectors, isolating the unit flows and opening drain valves to remove the water. If automatic response fails, the design permits 2 hours for operator action before water enters areas containing sodium piping. In the RCB and P,58, floor drain system leak detectors will signal leakage, allowing isolation of the affected water line. Upon confirmation of a sodium leak in area served by a recirculating gas cooling system,~the water and gas lines to the cooler serving the area will be valved off.
15.7.3.7 Sodium-Water Reaction in large Component Cleaning Vessel. The Large Component Cleaning Vessel (LCCV) is located in the Large Component Cleaning Cell in the lower part of the Reactor Containment Building. The event considered is a runaway reaction of sodium with water in the LCCV. Under normal operations, residual sodium or NaK on a component is removed by a controlled reaction with water vapor at 150*F in a nitrogen purged atmosphere; the reaction is controlled by regulation of l the water vapor concentration with the aim of maintaining hydrogen concentrations below 1 percent. A runaway reaction could occur by flooding the LCCV with water before the sodium was reacted with steam. The analysis considers cleaning the Intermediate Rotating Plug (a once-in-30-years task), and reacting water with 200 lb. of sodium on the bottom plate of the IRP; the resulting pressure in the LCCV is between 40 and 85 psig whereas it is designed for a static rupture pressure of 298 psig. The event assumes release of 100% of the radioactivity in the 200 lb. of sodium, decayed for 10 days. Such a release would isolate the RC8 and the effluent
; would pass through the filter system enroute to the environment, with decontamination factors of 20 for iodinc and 100 for particulates. The ; . applicants estimate maximum individual dose at the site boundary is 50 mrem to the thyroid.
The following summary of CRBR Design Basis Events / Systems Designs illustrates which systems are influenced by the design basis events. 37
~
SUMMARY
OF CRBR DESIGN BASIS EVENTS / SYSTEMS DESIGN I. Reactivity Insertion Events Operational Basis Earthquake , Safe Shutdown Earthquake Withdrawal at Startup , A. Withdrawal at Power Small Reactivity Insertions Loss of Hydraulic Holddown - Sudden Core Radial Movement Maloperation of Reactor Plant Controller B. Cold Sodium Insertion C. Gas Bubbles Passage Through Core Provides design bases for: l A. Primary and Secondary Reactor Shutdown Systems l Control Assembly Worth Speed of Response (Insertion)--Includes sensor trains / logic i
, (instrumentaticn and control systems) l Speed of Withdrawal :
Holddown System , Redundancy Diversity Reliability -
- 8. Cold Sodium Insertion--Provides design bases for administrative procedures, heat transport system instrumentation and control .
interlocks and PPS functions. , ! C. Gas Bubble Passage Through Core--The heat transport system design features to preclude gas bubbles from entering the core include: ' o Vents provided to eliminata possible gas pockets that may l form during sodium fill. L
, t o A low cover gas pressure which reduces gas entrainment.
o A continuous bleed from the top of the IHX is provided to '
- prevent accumulation of gas during operation. ;
38 , l l ; o . , ,_ __ _ . _ . _ _ _ -.
.-. . . - . - . . =-
l
SUMMARY
OF CRBR DESIGN BASIS EVENTS / SYSTEMS DESIGN (continued) I. Reactivity Insertion Events (continued) o The primary pump is designed to eliminate vortexing and gas entrainment. o .A high fluid velocity in the piping between the pump discharge and the vessel inlet minimizes the possiblility of
, . gas entrainment.
o A vortex suppressor is located in the outlet plenum at the optimum depth to prevent gas entrainment. o One or more holes with special pressure reducers will be 4 provided in the core support cone to vent gas from { i underneath the cone. i II. Undercooling Events A. Loss of Off-Site Electrical Power B. Spurious Pump Trips C. Inadvertent Closure of One Evaporator or Superheater Module Isolation Valve D. Turbine Trip E. Loss of Normal Feedwater j F. Inadvertent Actuation of the Sodium-Water Reactor Pressure Relief '
- System i
G. Pump Seizures H. Small Sodium to Water Leaks In Steam Generator Tubes I. Failure of the Steam Bypass System 4 J. Steam or Feedline Pipe Break K. Loss of Normal Shutdown Cooling System i ! L. Large Sodium-Water Reaction i M. Primary Heat Transport System Pipe Leak N. Intermediate Heat Transport System Pipe Leak 39
1
SUMMARY
OF CRBR DESIGN BASIS EVENTS / SYSTEMS DESIGN (continued) II. Undercooling Events (continued) Provides design bases for: A. Loss of Off-Site Electrical Power--Electrical power supply - systems, emergency power supply systems, auxiliary feedwater supply systems, steam generator auxiliary heat removal system and plant protection systems (including instrumentation and controls) - B. Spurious Pump Trip o Primary--Instrumentation and control system. o Intermediate--Instrumentation and control system. C. Inadvertent Closure of One Evaporator or Superheater Module isolation Valve--Instrumentation and control system, superneater safety valves, drum safety valves. D. Turbine Trip--Steam bypass system, instrumentation and control system. E. Loss Normal Feedwater--Instrumentation and control system, steam generator auxiliary heat removal system, superheater safety relief valves, auxiliary feedwater system. F. Inadvertent Actuation of the Sodium--Water Reaction Pressure Relief System--Instrumentation and control system, plant duty cycle for thermal stress effects on the steam generator modules, IHX, cold leg components and reactor vessel inlet nozzles. G. Pump Seizure o Primary Pump Seizure--Instrumentation and control systems, cold leg check valve. o Intermediate Loop Pump Seizure--Instrumentation and control system. H. Small Sodium-to-Water Leaks In Steam Generation Tubes--Instrumen-tation and control system, steam generator leak detection system, sodium drain system. I. Failure of tne Steam Bypass System--Instrumentation and control system, power relief valves, steam generator auxiliary heat removal system, auxiliary feedwater system. . I 40 l a- - - - - = e-->+
SUMMARY
OF CRBR' DESIGN BASIS EVENTS / SYSTEMS DESIGN (continued) II. Undercooling Events (continued) J. Steam or Feedline Break--Instrumentation and control system, plant duty cycle for thermal stress effects on components, superheater isolation system, safety valves, steam generctor auxiliary heat removal system, auxiliary feedwater system, steam generator building cell venting. K. Loss of Normal Shutdown Cooling System--Same as for I, above, failure of the steam bypass system. L. Large-Sodium-Water Reaction--Instrumentation and control system, leak detection system, sodium-water reaction pressure relief system, hydrogen separation system, steam generator isolation system. M. Primary Heat Transport System Pipe Leak--Leak detection system. N. Intermediate Heat Transport System Pipe Leak--Leak detection system ("not a design basis event at this time") III. Local Failure Events Provides design bases for: l 1 A. Fuel Assembly--Fission gas and tag gas detection system, delayed neutron monitoring system, fission product cleanup system. B. Control Assemblies--No specific pin failure detection system for control assemblies. I C. Radial Blanket Assembly--Tag gas and fission gas detection system, delayed neutron detection system. IV. _ Fuel Handling and Storage Events Provides design bases for: j A. Fuel Assembly Dropped Within Reactor Vessel During i Refueling--In-vessel transfer machine design and interlock features. B. Damage of Fuel Assembly Due to Attempt to Insert a Fuel Assembly Into An Occupied Position--Administrative procedures, IVTM grapple design features, fuel assembly outlet nozzle configuration features. i C. Single Fuel Assembly Cladding Failure and Subsecuent Fission-Gas Release During Refueling--Operating procedures, design features in the IVTM, grapples and EVTM and interlocks and controls. 1 I l 41 l
SUMMARY
OF CR'BR DESIGN BASIS EVENTS / SYSTEMS DESIGN (continued) 1 IV. Fuel Handling and Storage Events (continued) D. Cover Gas Release During Refueling--Design features, interlocks and controls and administrative procedure. E. The Heaviest Crane Load Impacts the Reactor Closure Head--Design features, operating and administrative procedures. F. Collision of EVTM with Control Rod Drive Mechanisms--Design teatures, operating procedures. V. Sodium Spills Provides design bases for: A. Primary Sodium In-Containment Storage Tank Failure During Maintenance--This event provides design basis pressure, temperatures and source terms for the reactor containment / confinement system. B. Failure of an Ex-Containment Primary Sodium Storage Tank--This event provides design basis pressure, temperatures and source terms for cell desi generator building)gn (in the intermediate bay of the steam C. Primary Heat Transport System Piping Leak--Provides pressure, temperature and source terms for design of reactor cavity and cell liners and cell structures, leak detection systems. D. Intermediate Heat Transport System Pipe Leak--Leak detection systems provides Jesign basis pressure, temperature and aerosol source term for steam generator building and cells. VI. Other Events Provides design bases for: A. Loss of D.C. System--Provides basis for design of redundant systems. B. Loss of Instrumentation or Valve Air--System designed to preclude loss of air to s6tety-related valves or instruments. C. IHX !eak--Leakage into primary system--sodium level sensing - devices in overflow tank and expansion tank. D. Off-Normal Cover Gas Pressure in the Reactor Coolant Boundary--Pressure regulation system redundancy, overpressure relief valves. 42
SUMMARY
OF CRBR DESIGN BASIS EVENTS / SYSTEMS DESIGN (continued) VI. Other Events (continued) E. Off-Normal Cover Gas Pressure in IHTS--Differential pressure measurement system, radiation monitoring system. 2 . F. Inadvertent Release of Oil Through Pumo Seals (PHTS)--Pump bearing and seal oil system, pump design features. G. Inadvertent Release of Oil Through Pump Seal (IHTS)--Same as for primary pump (F above). H. Generator Breaker Failure to Open at Turbine Trip--Generator and turbine trip logic. I. Rupture in the RAPS Cold Box--Radiation monitoring system, cell
- H and V vent system, cold box bypass system.
J. Liquid Radwaste System Failure (Leak or Rupture)--Design basis for cells and ventilation system. K. Failure (Leak or Rupture) in the EVST NaK System--Fire protection systems, catch pan design. L. Leakage From Sodium Cold Traps--Leak detection systems. M. Rupture in RAPS Noble Gas Storage Vessel Cell--Radiation monitoring system, cell vent system. N. Rupture in the CAPS Cold Box--Radiation monitoring system, cell H and V system. O. Leak In a Core Component Pot--Grapple load cells, periodic CCP inspection requirements, operating procedures.
! P. Spent Fuel Shipping Cask Drop From Maximum Possible Height--Cask handling procedures special rigging.
Q. Maximum Possible Conventional Fires, Flood, Storms or Minimum River Level--Plant fire protection systems, site drainage facilities. i R. Failure of Plug Seals and Annuli--Seal design, head access area radiation monitoring system. 3 S. Fuel Rod Leakage Combined With IHX and Steam Generator Leakage--intermediate cover gas pressure system, sodium leak l detection system. T. Sodium-Water Reaction in large Component Cleaning Vessel--Primary sodium removal and decontamination system, hydrogen detection system, vessel pressure and temperature design basis. 43 \ l
. - - _ - - _ _ _ = . _ _ . _ . _ . - . _ .- - -_- .-- _ _ -. . - -
SUMMARY
OF CRBR DESIGN BASIS EVENTS / SYSTEMS DESIGN (continued) DESIGN BASIS EVENTS / ACCIDENTS FOR ENGINEERED SAFETY FEATURES PSAR References 6.2.1 Reactor Confinement / Containment--Primary sodium tank failure, sodium . pool fire. 6.2.4 Containment Isolation System--Primary sodium tank failure, sodium - pool fire. 6.2.5 Annulus Filtration System--Site suitability source terms (major fission product release from the core). 6.2.6 Reactor Service Building (RSB) Filtration System--Site suitability source term. 6.2.7 Steam Generator Building Aerosol Release Mitigation System--Intermediate heat transport loop design basis leak. 6.3 Habitability Systems--Based on control room design source. 5.2 Reactor Guard Vessel--Reactor vessel leak, maintain sodium level at or above minimum safe level, i 5.3 Guard Vessels of PHTS Major Components--Maintain sodium level in reactor vessel at or above minimum safe level. 5.6 Residual Heat Removal System--Short and long term decay heat removal under normal and emergency conditions diverse, redundant, passive systems. 6.4 Cell Liner System--Design basis sodium leaks / spills. 6.5 Catch Pan and Fire Suppression Dash System--Non-radioactive liquid metal systems--prevent interactions with concrete. l l 44
i l D. COMPARISON OF CLINCH RIVER BREEDER REACTOR DESIGN BASIS ACCIDENTS WITH THOSE FOR OTHER LIQUID-METAL-COOLED FAST REACTORS The comparison of Clinch River Breeder Reactor (CRBR) design basis , accidents (DBAs) with those for other liquid-metal-cooled fast reactors is instructive because it provides a basis to evaluate the completeness of the
~
CRBR DBA delineation with respect to past practice and experience. t Based on the DBAs for the pre-CRBR fast reactors; EBR-II, SEFOR, FERMI [ and FFTF, it is clear that OBA delineation for CRBR generally reflects the l experience and engineering practice developed during the design, construction and operation of these reactors. Understandably, CRBR DBA delineation closely approximates that for the FFTF which is the latest i' precedent to CRBR. The Large Developmental Plant (LDP) which is planned for construction after CRBR also to a large extent, reflects DBA ; delineation developed from FFTF and CRBR experience. There are, however, ! several design features in the LDP which are not present in CRBR. One difference in the design of these reactors is the various means for providing for emergency shutdown heat removal. In the CRBR, emergency l shutdown heat removal is accomplished using the steam generator auxiliary , heat removal system (one of the three loops is adequate for this purpose). l An alternate system, Direct Heat Removal Service, is also included as a diverse, independent and functionally redundant system. In the EBR-II and , LDP designs the diverse, independent and functionally redundant alternate
! systems are in-vessel shutdown heat removal coolers with natural !
circulation capability (passive system). In FFTF and SEFOR, natural , circulation capability was provided as an alternate heat transport means in the normal and emergency decay heat removal systems. ! l l Another difference between the early plants i.e., EBR-II, SEFOR and ; FERMI and the later plants, FFTF, CRBR and LDP, is the provision in the early plants for primary containments specifically designed to accommodate ' the blast effects of TNT explosions presumed to be representative of highly J 45 '
- - . . , , , , - , - , - . . . , , y ---- ,. -.--- - - . - - -- - . - - - - - - , .- - --.--..,-% - ,, --. - ,-n..
__.m - _ i energetic hypotetical core disruptive accidents (HCDAs). In these plants (EBR-II, SEFOR and FERMI-1), HCDA energetics were estimated from modified l Bethe-Tait analyses. This highly simplified bounding analysis method i simply presumed beginning with a totally molten core whose upper half suddenly collapsed by gravity onto the lower half resulting in prompt criticality and hydrodynamic disassembly. In FFTF, CRBR and LDP, HCDA energetics are estimated using more mechanistic analytic models derived from extensive research which result in lower, more realistic estimated energy releases. l Core dispersal cones, (i.e., conical struct.res at the bottom of the reactor vessel designed to disperse molten fue; into a non-critical arrangement in the event of a core meltdown) were also a design feature in the early plants. Based on research which resulted in more detailed understanding of core behavior under hypothetical accident conditions and , the provision of independent, diverse, and functionally redundant reactor shutdown and decay heat removal systems, the FFTF designers concluded that f such a design feature was unnecessary. By virtue of the same logic they ' are not included in the CRBR and LDP designs. 1 A detailed comparison of the CRBR DBAs with those for the FFTF, SEFOR, EBR-II, LDP and FERMI plants is presented in the following sections.
- 1. Fast Flux Test Facility (FFTF)
Design basis accidents (DBAs) considered during the design of FFTF were reviewed against those for the Clinch River Breeder Reactor (CRBR) as a means of ascertaining the completeness of the CRBR DBA delineation. The review of DBAs for FFTF was based on the Final Safety Analysis Report (HEDL-TI-75-001), and the Safety Evaluation Report (NUREG-0358). I The DBA delineation for CRBR is quite similar to that for FFTF where , there are common features such as containment design bases and primary and intermediate heat transport piping and equipment cell design bases (sodj.um , fires). i 1 46 '
. - ~. .. . . . . ..
I l In FFTF the response of the reactor to extremel, enlikely accidents such as control assembly and single fuel assembly meltdown were analyzed as 1 reactivity' insertion accidents. They are not addressed in the CRBR PSAR. l t
- The toxicity effects of the release of sodium (oxide) aerosols were analyzed for FFTF. They are not addressed in the GRBR PSAR.
l The hypothetical core disruptive accident (150 MW-sec) is used in FFTF as a design basis for the reactor vessel, head, head bolts and vessel support arms to the extent that calculated plastic deformation of the j. components do not result in the expulsion of gross amounts of sodium and
- post accident heat removal can be sustaine'd.
Table 2 presents a comparison of the CRBR and FFTF design basis events. I l I 4 I 47
4 TAP,LE 2. COMPARISON Of FFTF DBAs WITil THOSE FOR CRBR CRBR Design Basis Events FFTF Design Basis Events I. 3eactivity Inseit_ ion Events O prational Basis Earthquake Not required for ffTF as reactor will be scranned at a peak ground acceleration 5 times less than that for a safe shutdown earthquake. Safe Shutdown Earthquake Safe Shutdown Earthquake analyzed. Withdrawal et Startup Withdrawal at startup analyzed. Withdrawal at Power Withdrawal at power analyzed. Small Reactivity Insertions Not addressed as a specific topic in the FSAR. Loss of flydraulic Holddown Loss of hydraulic holddown analyzed. g Sudden Core Radial Movement Sudden core rcdial movement analyzed. Maloperation of Reactor Plant N/A (Loop flow controller malfunctions analyzed). Controller Cold Sodium insertion Cold sodiunt insertion analyzed. r Gas Bubble Passage Through Core Gas bubble passage through core analyzed. Other Reactivity Insertion Events Analyzed. Control assembly meltdown. Single fuel assembly meltdown. Closed loop test section meltdown.
t , s . i TABLE 2. (continued) CRBR' Design Basis Events FFIF Design Basis Events - II. Undercooling Events Loss of Offsite Electrical Power loss of offsite power analyzed. Spurious Pump Trips Pump trips analyzed. Inadvertent Closure of One Evaporator N/A or Superheater Module Isolation Valve ! Turbine Trip N/A ] Loss of Normal feedwater N/A Inadvertent Actuation of the Sodium- N/A Water Reaction Pressure Relief System d; Pump Seizures Pump seizures analyzed.
- ' Small Sodium-to-Water Leaks in Steam N/A Gene.ator Tubes-i failure of the Steam Bypass System H/A Steam or feedline Pipe Break N/A Loss of Normal Shutdown Cooling Freezing of DilX sodium analyzed.
System 4 Large Sodium-Water Reaction N/A Primary lleat Transfer System Pipe Primary heat transport system leaks analyzed. Leak l 4
i 4 l TABLE 2. (continued) CRBR Design Basis Events FF1F Design Basis Events
- 11. Undercooling Events (continued)
Intermediate lleat Transport System Intermediate heat transport system leaks analyzed. Pipe Leak III. Local failure Events Fuel Assembly Fuel assenbly local failure events analyzed. Control Assenbly Control assembly local failure events analyzed. Radial Blanket Assenbly N/A (Stainless steel radial reflectors used). IV. Fuel llandling and Storage Events -
, fuel Asseubly Dropped Within Reactor In-vessel fuel handling accidents do not result in release of a Vessel During Refueling fission products.
Damage of fuel Asseubly due to Sensors in the in-vessel handling machine and discrimination Attempt to insert a fuel Assenhly on the fuel assembly inlet and produce such damage. into the Occupied Position Single Fuel Assembly Cladding Failure Protection against leakage of fission gases is provided and and Subsequent Fission Gas Release consequences of leakage are analyzed. During Refueling Cover Gas Release During Refueling Cover gas release during refueling analyzed. The lleaviest Crane Load Impacts the Load drops analyzed. Ileavy loads lifted over operating deck Reactor Closure llead prohibited administrative 1y. D g 9
T ABL E 2. (continued) 4 CRBR Design Basis Events FFTF Design Basis Events l IV. Fuel llandling and Storage Events (continued) Collision of EVTM with Control Rod N/A (Control rod drives are below operating deck). Drive Mechanisms V. Sodium Spills Primary Sodium In-Containment Storage 500 lb Na atomized as result of IICDA results in 3.0 psig
, Tank failure During Maintenance pressure.
25,000 lb Na burned results in 1.7 psig. i Leaks in primary heat transpcrt system cells form basis for containment and cells design pressure and temperatures. Failure of an Ex-Containment Primary N/A Sodium Storage Tank Primary lleat Transport System Piping Effect of leaks on containment and cells pressures and Leak s temperatures analyzed.
- Intermediate lleat Transport System Analyzed for cell pressure and temperature design basis.
pipe Leak Other Analyses Sodium toxicity limits analyzed VI. Other Events Loss of DC System Analyzed. i i
i T ABL E 2. (continued) CRBR Design Basis Events FFTF Design Basis Esents '
- VI. Other Events (continued)
Loss of Instrumentation or Valve Air Analyzed. lilX Leak tilX leak analyzed (Secondary side pressure higher than primary l side). Of f-Nonnal Cover Gas Pressure in the Analyzed Reactor Coolant Boundary , Of f-normal Cover Gas Pressure in IHTS Analyzed inadvertent Release of Oil Through The primary sodium pump design provides safeguards a ainst Pump Seal (PitTS) ingress of seal lubricant to the top of the pump tan . i l" Inadvertent Release of Oil Through provid gainst Pump Seal (IllTS) The secondary sodium pump desiingress of seal lubricant to tEie top ofs safeguardsthe pump tank. Generator Breaker Failure to Open N/A
- at Turbine Trip Rupture in the RAPS Cold Box Analyzed
! Liquid Radwaste System f ailure (Leak Analyzed or Rupture) l i l .I 1 e e a j e 4
. . + ,
l TABLE 2. (ca..t inued) 4 CRBR Design Basis Events FFIF Design Basis Events i VI. Other Events (continued) Failure (Leak or Rupture) in the EVST Analyzed for interim decay storage. NaK System Leakage from Sodium Cold Traps Analyzed
; Rupture in RAPS Noble Gas Storage Analyzed Vessel Cell Rupture in the CAPS Cold Box Analyzed ,
Leak in a Core Component Pot Analyzed Spent Fuel Shipping Cask Drop from N/A--no fuel shipping cask handling in containment. Maximum Possible lleight 10 Maximum Possible Conventional Fires, Considered in design of plant. Flood, Storms or Minimum River Level i Failure of Plug Seals and Annuli Plugs and seals were analyzed. I fuel Rod Leakage Combined with I lix N/A and Steam Generator Leakage Sodium-Water Reaction in large Not explicitly addressed in FFIF FSAR. Sodium removal system Component Cleaning vessel in IEM cell designed to limit quantity of water in system. i t i I l
_ _ _ . _ - _ . _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ = . - - - _m _.. _ _ ___ _ ._ _ _ _ _ m T ABL E 2. (continued) Design Basis Events / Accidents for Engineered Safety features FFTF Design Basis Events i Reactor Confinement / Containment "The containment design shall be adequate to sustain (1) a 1 hypothetical event which may cause disruptive disassembly of the reactor core with accompanying energy release and decay heat, and (2) the maximum postulated radioactive sodium fire." Containment only. Design basis is sodium fire from IICDA. i Containment isolation System CIS included in FFTF. 1 Annulus filtration System N/A Reactor Service Building (RSB) Filtration N/A System i
- Steam Generator Building Aerosol Release N/A o Mitigation System s
IIabitability System Design considered llCDA for source of aerosols and radiation. i Reactor Guard Vessel included in design. Guard Vessel of PitTS Major Components included in design. Residual lleat Removal System Natural circulation. Cell Liner System included in design. Catch Pan and Fire Suppression Deck Included in design.
- System 4
4 . , f
- 2. Experimental Breeder Reactor-II (EBR-II) l Design basis accidents (DBAs) considered during the design of EBR-II l
- were reviewed against those for the Clinch River Breeder Reactor (CRBR) as 4
one means of ascertaining the completeness of the CRBR DBAs. The review of DBAs for EBR-II was based on ANL-5719, Summary Report of 1 - the Hazards of the EBR-II, May 1957 and ANL-5719 (Addendum), Addendum to Hazard Summary Report Experimental Breeder Reactor-II (EBR-II), June 1962. The review was consciously restricted to these two documents on the basis that they best represented the stah of the art in fast reactor design in the U.S. at that period in time (1957-1962). It is known that many reanalyses of the EBR-II under abnormal and severe accident conditions have been carried out to support the changing role of the EBR-II in the nation's fast reactor development program and that these analyses which employ'more refined analytic techniques have demonstrated the extreme conservatisms a embodied in the initial design basis accident assumptions. I i As mentioned earlier (pg. 45), the EBR-II design incorporated in-vessel shutdown heat removal coolers. This natural circulation heat removal system provided the requisite diverse, independent and functionally redundant decay heat removal system. The consideration of containment and filtering of sodium oxide aerosols from sodium fires was not specifically addressed in the EBR-II ,
, Hazard Summary Report. They are considered in CRBR design.
Sodium-concrete reactions were not explicitly addressed in the EBR-II I Hazard Summary Report for either reactor building containment design (pressures, temperatures, isolation and venting) or boiler plant building design. They are considered in CRBR design. Fuel element bowing was a major design consideration in EBR-II as a result of the EBR-I reactivity instability due to fuel element bowing. These design considerations did not include the neutron-radiation-stainless 55
steel swelling phenomenon (it was not discovered until 1967). Stainless steel swelling is a major design consideration in CRBR core design. Design of the EBR-II primary and secondary containment systems were based on containing a highly energetic hypothetical core disruptive accident (HCDA). The energy release was estimated using a Bethe-Tait type ~ analysis. In CRBR, by virtue of the two independent, diverse and functionally redundant reactor shutdown and decay heat removal 3ystems, the HCDA is deemed so unlikely that it need not be included as a design basis accident and hence is not addressed in the Preliminary Safety Analysis Report. However, response of the plant to more mechanistically-derived HCDAs using advanced accident analysis codes (such as SIMMER) is studied to determine structural and thermal margins beyond the design bas"s (SMBDB and TMBDB,respectively). Table 3 presents a comparison of the CRBR DBAs with those discussed in the EBR-II Hazard Summary Reports. The table indicates that CRBR accident delineation is more extensive than that for EBR-II. 4 I 56 w e-.. .
l l 6 TABLE 3. COMPARISON OF EBR-Il DBAs WITil Til0SE FOR CRBR CRBR Design Basis Events EBR-Il Design Basis Events
- 1. Reactivity insertion Events Operational Basis Earthquake '
In the CRBR, the OBE and SSE events are analyzed as step reactivity insertions of 30t and 60s, respectively. Similar I analyses for EBR-il were not explicitly addressed in the Safe Shutdown Earthquake llazards Sunnary Reports. EBR-il is scranned if an earthquake s is sensed. Withdrawal at Startup " . . . Six hypothetical cases were examined to assess the safety characteristics of the reactor and to estimate the Withdrawal at Power consequences." Small Reactivity insertions "The submerged design and coolant pump inlet features renders impossible the rapid addition of moderator to the core via the reactor cooling system." Fuel element bowing as a source of m positive reactivity addition was also addressed in detail as a result of the accident in EBR-1. Loss of Ilydraulic flo1down Mechanical stops incorporated in top cover. 4 Sudden Core Radial Movement flot explicitly addressed in the llazards Sunnary Reports. Maloperation of Reactor Plant Malfunctions of the automatic flux level control system were Controller analyzed. Cold Sodium Insertion "Further, the innense volume (and ef fective heat capacity) of bulk sodium from which the reactor inlet coolant is drawn renders impossible any sudden changes in coolant inlet temperature with its attendant reactivity changes." 1 i 1
TABLE 3. (continued) CRBR Design Basis Events EBR-il Design Basis Events Gas Bubble Passage Through Core "The submerged design and coolant pump inlet features renders impossible the rapid addition of moderator to the core via the reactor cooling system."
- 11. Undercooling Events Loss of Offsite Electrical Power Three emergency diesel generators provided as well as battery power for an auxiliary electromagnetic pump for shutdown heat removal. In-tank shutdown coolers using natural circulation diso available for shutdown heat removas.
Spurious Pump Trips Given scram, the reactor can tolerate loss of all pumping power without daaage. I Inadvertent Closure of One Evaporator Malfunctions of the evaporators and superheaters were not or Superheater Module Isolation Valve explicitly addressed in the llazards Summary Report. Given the g; design of the reactor primary, secondary and steam generation systems one can presume that EBR-il is as insensitive to such malfunctions as CRBR. ! Turbine Trip Not explicitly analyzed in the llazards Summary Reports. Loss of Normal Feedwater Not explicitly analyzed in the llazards Summary Reports. Inadvertent Actuation of the Not explicitly analyzed in the llazards Sunnary Reports. Sodium-Water Reaction Pressure Relief System Pump Seizures Given scran, the reactor can tolerate loss of all pumps without damage.
T ABL E 3. (continued) CRBR Design Basis Events EBR-il Design Basis Events Small Sodium-to-Water Leaks in Steam Not explicitly addressed in the llazards Sunnary Reports. Generator Tubes failure of the Steam Bypass System Not explicitly addressed in the llazards Sunnary Reports. Steam of Feedline Pipe Break Not explicitly addressed in the llazards Sunnary Reports. Loss of Normal Shutdown Cooling System Passive, natural circulation, in-tank shutdown coolers installed. Large Sodium-Water Reaction Separate building for steam generators. Building has
" blowout" panels. Sodium relief system provides for pressure surge acconnodation. No discussion of aerosols.
Primary tieat Transport System Pipe All primary system piping is contained within the primary m Leak tank. "The triple tank features and the sodium system and the submerged design concept removes the loss of coolant accident
- from the list".
Intermediate lleat Transport System Provisions for the acconnodation of intermediate system pipe 1 Pipe Leak leaks are not explicitly addressed in the Ilizards Sunnary Reports. Ill. Local Failure Events Fuel Assembly Fuel element failure propagation is discussed at length in the llazard Sunnary Reports. Control Assembly in EBR-Il the control rods use fuel elements for reactivity control hence, discussion of local failure events is the same as for the fuel assenbly. 2 I t
- . - _ . __ . . .. - -- - - _ _ _ = -
TABLE 3. (continued) CRBR Design Basis Events EBR-Il Design Basis Events Radial Blanket Assembly Analysis of local failure events in radial blanket assemblies is not explicitly addressed in the llazard Sunnary Reports. j IV. Fuel llandling and Storage Events 1 Fuel Assembly Dropped Within Reactor fuel handling accidents were considered in the design of the Vessel During Refueling systems. A " subassembly basin" was incorporated within the primary tank to position a dropped assembly such that it could be cooled by natural convection. Dainage of Fuel Assembly Due to Not explicitly addressed in the llazard Sunnary Reports. Attempt to Insert a fuel Assembly into An Occupied Position Single fuel Assembly Cladding failure Not explicitly addressed in the llazard Sununary Reports. and Subsequent Fission Gas Release
$ During Refueling
, Cover Gas Release During Refueling Not explicitly addressed in the llazard Summary Reports. The lleaviest Crane Load Impacts the Not explicitly addressed in llazard Sunnary Reports. , Reactor Closure llead Collision of Evil 4 With Control Rod EBR-Il does not employ an ex-vessel transfer machine. Drive hechanisms 4 f 8 . 8 e
TABLE 3. (continued) CRBR Design Basis Events EBR-II Design Basis Events V. Sodium Spills Primary Sodium In-Containment Storage Tank Failure During Maintenance These events per se are not explicitly discussed in the llazard Sununary Reports. The reactor building contairunent design is failure of An Ex-Containment Primary based on pressures arising from sodium-air reaction. Sodium Storage Tank Primary lleat Transport System Piping All the primary piping is contained within the primary tank. Leak s The primary tank and the reactor containment building are designed to acconmodate the effect of a hypothetical core disrupture accident. Intermediate lleat Transport System Not explicitly addresscJ in the llazard Sunnary Reports. Pipe Leak VI. Other Events Loss of DC System State-of-the art " fall safe" instrument and control systems Loss of Instrumentation or Valve Air were incorporated in the design. lilX teak IllX operating pressure is higher than that for the primary system--leakage would be into the primary system (as in CRBR). The lilX is in the primary tar:k. Off-Normal Cover Gas Pressure in the Not explicitly addressed in the llazard Sunnary Reports. Reactor Coolant Boundary J
l 1 l l l TABLE 3. (continued) ' l CRBR Design Basis Events EBR-Il Design Basis Events j Of f-Normal Cover Gas Pressure in 7 4tTS Inadvertent Release of 011 Through Pump Seal (PiliS) Inadvertent Release of Oil Through Pump Seal (IIITS) Ger,eg;3r Breaker Failure to Open dl Ihauihe Irip Rupture in the RAPS Cold Box )Theseeventswerenotexplicitlyaddressedinthellazards Suonary Reports. Liquid Radwaste System Failure g (Leak or Rupture) Failure (Leak or Rupture) in the EVST NaK System Leakage From Sodium Cold Traps Rupture in RAPS Noble Gas Storage Vessel Cell I Rupture in the CAPS Cold Box Leak In a Core Component Pot , e e e e
.-. . . - . . . . _ . ___ _ - . . -. _ - -. . - ~ _ _ _ . -
T ABL E 3. (continued) CRBR Design Basis Events EBR-II Design Basis Events Spent fuel Shipping Cask Drop from Not explicitly addressed in llazard Sunnary Reports. Maximum Possible lleight Maximum Possible Conventional Fires, Not discussed in any detail in the llazard Sunenary Reports. Flood, Storms of Minimum River Level f ailure of Plug Seals and Annuli Analyzed as part .f the analysis for acconnodation of hypothetical core disruptive accidents.
< Fuel Rod Leakage Combined With lilX Not explicitly addressed in Hazard Sunnary Reports.
and Steam Generator Leakage Sodium-Water P.? action in large EBR-II does not employ a large component cleaning vessel.
- Component Cleaning Vessel Design Basis Events / Accidents for Engineered Safety Features Reactor Confinement / Containment The single shell reactor building containment is designed to i
acconnodate the pressures and temperatures resulting f rom a design basis sodium fire and/or a significant nuclear incident. Containment Isolation System Containment isolation is based on sensing a significant nuclear incident or primary sodium fire. Annulus Filtration System A single shell reactor building containment is used in EBR-II. Reactor Service Building (RSB) Not explicitly addressed in Hazard Sunnary Reports. Filtration System 4 i i I
TABLE 3. (continued)
- CRBR Design Basis Events EBR-Il Design Basis Events ,
, Steam Generator Building Aerosol Not explicitly addressed in the llazard Sunnary Reports.
- Release Mitigation System
, liabitability Systems Not explicitly addressed in the llazard Sunnary Reports. 1 l Reactor Guard Vessel in EBR-il the reactor vessel is continued within a
- double-walled primary tank surrounded by a blast shield i surroJnded by the biological shield. The design basis for the primary containment system was. the acconnodation of a hypothetical energetic core disruptive accident.
Guard Vessels of PitTS Major All PilTS components are contained within the primary tank. Components Residual lleat Removal System The residual heat removal systems emphasize natural circulation capability. The in-vessel shutdown coolers , provide a virtually passive, natural circulation system for decay heat removal. Cell Liner System Provision for prevention of sodium-concrete interactions are
?? not explicitly addressed in the llazard Sunnary Reports.
Catch 'an and Fire Suppression Provision for prevention of sodium-concrete interactions are Deck 5. 2m not explicitly addressed in the llazard Sunanary Reports. 6 6 0
- g 9
- 3. Southwest Experimental Fast 0xide Reactor (SEFOR)
Design basis accidents (DBAs) considered during the design of SEFOR were reviewed against those for the Clinch River Breeder Reactor (CRBR) as a means of ascertaining the completeness of the CRBR DBA delineation. The review of DBAs for SEFOR was based on the Facility Description and
~
Safety Analysis Report. 1 The SEFOR plant design is quite different than that for CRBR principally in that SEFOR employed an inerted refueling cell directly over the reactor vessel head as opposed to the in-vessel fuel handling system in CRBR. Additionally, the primary containment, a massive reinforced-concrete structure, was designed to accommodate a maximum hypothetical accident equivalent to 200 pounds of TNT. Further, SEFOR was not designed to produce electricity so the heat was discharged to the atmosphere via air blast heat exchangers. i . In addition to auxiliary pumping systems for shut-down heat removal the reactor plant was designed for natural circulation as an additional means for decay heat removal. Leak-bef6re-break criteria were employed in the analyses for effects j of coolant leaks. The fuel element coolant inlet design would not have precluded blockage (as occurred in FERMI-1). However, celtdown of a fuel assembly by blockage was analyzed and it was concluded that the reactivity effects were
- within the shutdown system capability and that failure propagation (assembly-to-assembly) would not' occur.
A core dispersion assembly was installed directly below the reactor i vessel. Table 4 presents a comparisen of CRBR and SEFOR design basis events. l i 65
i TABtE 4. COMPARISON Of SEFOR DBAs WITH THOSE FOR CRBR CRBR Design Sasis Events SEf 0R Design Basis Events I. Reactivity Insertion Events Operational Basis Earthquate Reactor scrammed at 1/6th of design basis earthquake intensity. Saf e Shutdown Earthquake Design basis earthquake used in design of structures and critical components. Withdrawal at Startup Numerous reactivity insertion events were analyzed including events associated with the oscillator and the fa-t reactivity Withdrawal at Power excursion device. The reactor employed radial reflectors (rather than control rods) as the reactisity control method. Snall Reactivity insertions Because the reactor was designed to quantify the transient response of a fast, mixed-oxide reactor, extensive analyses of reactivity coefficients were performed. Fuel element bowing effects were addressed in detail in design of the fuel elements. i f oss of Ilydraulic llolddown flydraulic holddown not employed. Effect of axial fuel 2 displacement analyzed. Sudden Core Radial Movement the core was clamped in a minimum diameter condition, i.e., radial compaction of the core was not possible. Maloperation of Reactor plant Numeous safety features were employed to protect the reactor Controller from malfunctions of the oscillator control systems. Cold Sodium insertion Cold sodium insertion analyzed. Gas Bubble Passage Through Core Gas bubble passage through core analyzed. II. tindercooling Events Loss of Of f site Electrical power Loss of electrical power analyzed. Natural circulation employed for shutdown heat removal as backup to secondary and auxiliary cooling systems. Spurious pump Trips Loss of pumping addressed in the design and redundant pumps and power supplies incorporated. Inadvertent Clnsure of One Evaporator N/A or Superheater Module Isolation Valve Turbine Trip N/A Loss of Normal Feedwater N/A
. - _ _ _ - - _ _ . - . _ - _ _ _ _ _ _ - = _ _ _ . _ _ _ - ,-
j . T ABLE 4. (continued) CRBR Design Basis Events SEFOR Design Basis Events __ II. Undercooling Events (continued) ! Inadvertent Actuation of the Sodium- N/A Water Reaction Pressure Relief , System 4 Pump Seizures N/A to electromagnetic pumps. , Small Sodium-to-Water Leaks in N/A } Steam Generator Tubes Failure of the Steam Bypass System N/A ! Steam or Feedline Pipe Break N/A i Loss of Normal Shutdown Cooling Auxiliary coolant system employed. Natural circulation l g System cooling for shutdown heat removal available. Large Sodium-Water Reaction N/A Primary Heat Transport System Pipe Leak-before-break criterion invoked. Analyses for double-Leak double-ended pipe breaks performed. Shutdown heat removal I through primary, auxiliary or makeup system available. , Intermediate lieat Transport System 2 Pipe Leak s III. Local Failure Events i i Fuel Assen61y Fuel element coolant inlet design would not have precluded a 1 blockage as occurred in Fermi-I. Meltdown and a fuel element i by blockage was analyzed and it was concluded that the I reactivity effects were within the control system capability and that failure propagation Would not occur. Control Assembly The principal reactivity control system employed radial 4 reflectors around the reactor vesssel. In-core shim rods (84C) were also used. l Radial Blanket Assembly N/A IV. Fuel Handling and Storage Events 1 Fuel Assembly Dropped Within Reactor The SEFOR employed a ref ueling cell directly over the reactor. Vessel During Refueling for refueling, the reactor vessel head was removed, exposing the complete core. Reactivity effects and dropping a fuel { assembly into a core position were analyzed. 1 l
TABLE 4. (rontinued) , CRBR Design Basis Events SEFOR Design Basis Events IV. Fuel llandling and Storage Events Icontinued) Damage of Fuel Assembly Due to Precluded by visibility of tops of fuel elements during i Atteinpt to insert a fuel Assembly refueling. into an Occupied Position
- Single fuel Assembly Cladding failure lhe SEf 0R fuel was not subjected to high burnup. the refueling and Subsequent Fission-Gas Release cell was part of the primary containment which was designed to During Refueling accoimnodate fission product releases attendant with an llCDA.
l Cover Gas Release During Refueling The SEFOR employed a refueling cell directly over the reactor. For refueling, the reactor vessel head was removed exposing the complete core. The refueling cell is a part of the primary containment. The lleaviest Crane Load Impacts the Movement of the reactor inner head between $?s:el and storage Reactor Closure Head involves use of the multiton grapple. The reactor inner head maximum potential energy and configuration are such that it is not possible to drop it into the reactor vessel to cause core compaction. co Collision of EVIH with Control Rod Not applicable i Drive Mechanisms V. Sodium Spills Primary Sodium In-Containment Storage Major hypothetical sodium spills were analyzed and it was shown . Tank failure During Maintenance that the secondary containment shell would not be overpressurized. Sodium-concrete interactions were not failure of an Ex-Containment Primary explicitly addressed in the Safety Analysis Report. The outer Sodium Storage Tank containment had a design pressure of 30 psig. Primary lleat Transport System Piping Leaks Intermediate lleat Transport System Pipe Leak VI. Other Events Loss of DC System Analyses of faults in the DC system were not explicitly addressed in the Safety Analysis Report. Battery power was provided as a backup DC power supply. , Loss of Instrumentation of Valve Ai: Analyses of faults in the instrument air supply system were , e not explicitly addressed in the Safety Analysis Report. u f
~-- _ __ _ - - _ - - . _ _ _ _ _ _ _ - . -
IABLE 4. (continued) CRBR Design Basis Events Sff 0R Design Basis Events VI. Other Events (continued) lilX Leak The secondary side of the IHXs are at a higher pressure than the primary side so leakage would be into the primary system. > Of f-Normal Cover Gas Pressure in the Detailed analyses of cover gas system faults were not Reactor Coolant Boundary explicitly addressed in the Safety Analysis Report but the report discusses provisions for overpressurization protection. Off-Normal Cover Gas Pressure in IlifS Inadvertent Release Oil through Pump N/A--electromagnetic pumps. Seal (PHIS) Inadvertent Release of Oil Through N/A--electromagnetic pumps. Pump Seal (IHIS) , Generator fireaker Failure to Open N/A dt Turbine Trip Rupture in the RAPS Cold Box A cold box was not employed in SEFOR. Argon purification in the refueling cell was accomplished using a NaK bubbler. The cover gas was a feed and bleed system. The bleed was to a E waste gas header and then to a storage tank for later discharge. Detailed analyses of faults in the NaK-bubbler system were not explicitly addressed in the Safety Analysis Report. Liquid Radwaste System failure Detailed analysis of failures in the liquid radwaste system (Leak or Rupture) were not explictly addressed in the Safety Analysis Report. , failure (Leak or Rupture) in the EYST trradiated fuel storage tank leaks and heat removal faults NaK System were analyzed. Leakage From Sodium Cold Traps Detailed analyses of faults in the sodium cold trap systems were not explictly addressed in the Safety Analysis Report. Rupture in RAPS Noble Gas Storage Detailed analyses of rupture of the radioactive argon storage Vessel Cell tanks were not explictly addressed in the Safety Analysis Report. Rupture in the CAPS Cold Box A cold box was not employed in SEFOR. Leak in a Core Component Pot N/A Spent Fuel Shipping Cask Drop from Shipping cask drop accidents were not explicitly addressed in , Maximum Possible Height the Safety Analysis Report. Maxinuun Possible Conventional Fires Plant tornado hardened for 300 mph wir's. Flood, Storms or Minimum River Level
! __ _ (, _ _ _- --
TABLE 4. (continued) CRBR Design Basis Events SEFOR Design Basis Events i VI. Other Events (continued) Failure of plug Seals and Annull Detailed analyses of failure of plug seals and annult were not explicitly addressed in the Safety Analysis Report. Fuel Rod Leakage Combined with lilX N/A--no steam generators in SEFOR. and Steam Generator Leakage Sodium-Water Reaction in large N/A Component Cleaning Vessel i i Design Basis Events / Accidents for Engineered Safety features SEFOR Design Basis Events Reactor Confinement / Containment The inner or primary containment structure was designed to acconenodate an energetic hypothetical core disrupture accident. An outer Containment of steen was also inciuded in the design. Containment Isolation System Containment isolation capability was provided. Annulus Filtration System An annulus filtration system was not included in the SEFOR design. Reactor Service Building (RSB) Filtration N/A for SEFOR. System Steam Generator Building Aerosol Release N/A Mitigation System liabitability Systems Control room shielding and heating and ventilating under t accident conditions were addressed in the design of SEFOR. Reactor Guard vessel Guard vessels were provided in the SEFOR design. i Guard Vessels of PliTS Major Components Guard vessels were provided in the SEFOR design. 1 j Residual ifeat Removal System Auxiliary decay heat removal systems were included in the 4 SEFOR design. Cell Liner System I Cell liners were incorporated in the SEFOR design but to prevent in-leakage of air rather than prevent sodium-concrete interactions. Catch pan and Fire Suppression Deck System Detailed discussion of catch pan and fire suppression decks
- were not presented in the Safety Analys is Report.
I TABLE 4. (continued) , Other Engineered Safety Features In SEFOR Blast shielding--Blast shielding was provided to protect the inner containment from the blast effects of a 200 pound INT explosion at the center location of the reactor core. Core Dispersion Assemblies--Core dispersion assemblies were provided to reduce the probability of reassembly of the reactor fuel into a critical mass, if disassembly of the core and reactor should occur. The core dispersion assembly was installed directly below the reactor vessel. 23 !
+- __ _ __.
I
- 4. LMFBR Conceptual Design Study--Large Developmental Plant (LDP)
Design basis accidents (DBAs) considered during the design of the Large Developmental Plant (LDP) were reviewed against those for the Clinch River Breeder Reactor (CRBR) as one means of ascertaining the completeness of the CRBR DBA delineation. ' The review of DBAs for the LDP was based on CDS 400-7, LMFBR Conceptual Design, Design Safety Review Report, Revision 1, September 30, 1981. In general, the DBAs considered in the LDP follow directly from those used in FFTF and CRBR except in those instances where design changes eliminated particular components or systems (such as the RAPS Cold Box) and hence the need for analysis of the radiological consequences resulting from failure of the system or component. A difference between the LDP and CRBR is the provision of four heat transport loops in the LPD (CRBR has three) because, as stated in CDS 400-7, " Loss of piping integrity accident issues require consideration of use of more than three loops or alternatives to protect the core from damage. Decay heat removal reliability is generally improved with larger numbers of primary loops if the remainder of the decay heat removal system l is unchanged." Another difference between the LDP and CRBR design is the provision in the LDP design for an additional completely passive, safety-grade, l independent and diverse system for shutdown heat removal; the Direct Reactor Auxiliary Cooling System (DRACS). The concept is similar to that employed in EBR-II (i.e., in-vessel shutdown heat removal coolers). Table 5 presents a comparison of CRBR and LDP design basis events. ' 1 72 I
TABLE 5. COMPARISON OF LDP DBAs WITH THOSE FOR CRBR CRBR Design Basis Events LDP Design Basis Events I. Reactivity Insertion Events Operational Basis Earthquake The OBE and SSE are addressed in the Safe Shutdown Earthquake LDP design. Withdrawal at Startup Withdrawal at control speed analyzed Witndrawal at Power as an anticipated event. Withdrawal at maximum mechanical speed analyzed as extremely unlikely event. Small Reactivity Insertions Analyzed as anticipated events Loss of Hydraulic Holddown Analyzed as unlikely event Sudden Core Radial Movement Analyzed as unlikely event Maloperation of Reactor Plant Analyzed as unlikely event Controller Cold Sodium Insertion Analyzed as extremely unlikely event Gas Bubble Passage Through Core Analyzed as extremely unlikely event Other Reactivity Insertion Events Drop of single control rod at full power analyzed as anticipated event. II. Undercooling Events Loss of Offsite Electrical Analyzed as an anticipated event. Power Transition to natural circulation cooling does not result in unacceptable cladding temperatures. Spurious Pump Trips Analyzed as an anticipated event Inadvertent Closure of One Analyzed as an anticipated event Evaporator or Superheater Module Isolation Valve Turbine Trip Analyzed as an anticipated event Loss of Normal Feedwater Analyzed as an anticipated event Inadvertent Actuation of the Analyzed as an anticipated event Sodium-Water Reaction Pressure Relief System Pump Seizures Analyzed as unlikely event 73
TABLE 5. (continued) CRBR Design Basis Events LDP Design Basis Events II. Undercooling Events (continued) Small Sodium-to-Water Leaks in Protection from small sodium-water . Steam Generator Tubes reactions is provided by rupture disks in the gas equalization line from the IHTS expansion tanks to - the SWRP tanks. Failure of the Steam Bypass Analyzed as unlikely event System Steam or Feedline Pipe Break Analyzed as unlikely event Loss of Normal Shutdown Cooling Two diverse shutdown heat removal System systems available--SGACS and DRACS Large Sodium-Water Reaction A sodium-water reaction pressure relief subsystem is provided Primary Heat' Transport System Leak-before-break philosophy used. Pipe Leak Leak detections employed. Cells ! are steel-lined to prevent Na-concrete interactions. Analyzed as extremely unlikely event. Intermediate Heat Transport Analyzed as extremely unlikely event System Pipe Leak III. Local Failure Events Fuel Assembly Local failure events include sto-Control Assembly chastic pin failures, overenrich-Radial Blanket Assembly ment / overpower errors, local flow blockages, postulated inlet blockages and the passage of small , gas bubbles. These events have been studied extensively for all assemblies . . . . All results indicate that an isolated nin failure will neither fail adjacent pins nor damage the duct in which it resides and that other local - fault initiators do not represenc potential hazards. , 74
TABLE 5. (continued) CRBR Design Basis Events LDP Design Basis Events IV. Fuel Handling and Storage Events Fuel Assembly Dropped Within Analyzed as unlikely event Reactor Vessel During , Refueling Damage of Fuel Assembly Due to Analyzed as unlikely event Attempt to Insert a Fuel Assembly into an Occupied Position S'ngle Fuel Assembly Cladding Analyzed as unlikely event Failure and Subsequent Fission Gas Release During Refueling Cover Gas Release During Analyzed as unlikely event Refueling The Heaviest Crane Load Impacts -- the Reactor Closure Head Collision of EVTM with Control There is no EVTM in the LDP Rod Drive Mechanisms Other Events Loss of offsite power during refueling--analyzed as an anticipated event. Failure to seat assembly properly. Jamming or malfunction of fuel transport system. V. Sodium Spills Primary Sodium In-Containment Analyzed as extremely unlikely event Storage Tank Failure During Maintenance Failure of an Ex-Containment Not explicitly addressed in Design Primary Sodium Storage Tank Safety Review Report Primary Heat Transport System Analyzed as extremely unlikely I Piping Leaks events. Provides design bases for o cell pressures and temperatures. s 75
TABLE 5. (continued) CRBR Design Basis Events LDP Design Basis Events V. Sodium Spills (continued) Intermediate Heat Transport Analyzed as extremely unlikely , System Pipe Leak (vents. Complete severance of 36-inch pipe evaluated. Sodium aerosols retained in SGB. - Other Events Sodium spill in the fuel transfer cell due to maintenance activities. Failure of EVST sodium cooling system during operation. VI. Other Events 1 Loss of DC System Analyzed as anticipated event Loss of Instrumentation or Analyzed as anticipated event Valve Air IHX Leak IHX leaks were considered in the LDP. The IHX sodium is at a greater pressure than the primary sodium so leakage would be into the primary system. Off-Normal Cover Gas Pressure Analyzed as anticipated event in the Reactor Coolant Boundary Off-Normal Cover Gas Pressure Analyzed as anticipated event in IHTS Inadvertent Release of Oil Analyzed as unlikely event Through Pump Seal (PHTS) Inadvertent Release of Oil Not explicitly addressed in Design Through Pump Seal (IHTS) Safety Review Report Generator Breaker Failure to Not explicitly addressed in Design Open at Turbine Trip Safety Review Report Rupture in the RAPS Cold Box Not explicitly addressed in Design Safety Review Report Liquid Radwaste System Failure Analyzed as unlikely event (Leak or Rupture) 76 1
TABLE 5. (continued) [ CRBR Design Basis Events LDP Design Basis Events l VI. Other Events (continued) Failure (Leak or Rupture) in Not explicitly addressed in Design the EVST NaK System Safety Review Report
. Leakage frcm Sodium Cold Failure of cold traps and cesium Traps traps analyzed Rupture in RAPS tjoble Gas Analyzed as unlikely event Storage Vessel Cell Rupture in the CAPS Cold Box Not explicitly addressed in Design Safety Review Report Leak in a Core Component Pot Not explicitly addressed in Design Safety Review Report Spent Fuel Shipping Cask Drop Analyzed as extremely unlikely event from Maximum Possible Height (during refueling)
Maximum Possible Conventional Not explicitly addressed in Design Fires, Flood, Storms or Minimum Safety Review Report River Level Failure of Plug Seals and Analyzed as extremely unlikely event Annuli Fuel Rod Leakage Combined with Not explicitly addressed in Design IHX and Steam Generator Leakage Safety Revies Report Sodium-Water Reaction in Large Not explicitly addressed in Design Component Cleaning Vessel Safety Review Report VII. Design Basis Events / Accidents f or Engineered Safety Features Reactor Confinement / Containment The LDP incorporates a containment / confinement system. The containment DBA is a sodium fire resulting from a PHTS pipe leak during maintenance while the PHTS ' cell is open to the containment. Containment Isola: ion System The LDP incorporates a containment isolation system Annulus Filtration System The LDP incorporates an annulus isolation system 77
TABLE 5. (continued) . CRBR Design Basis Events LDP Design Basis Events VII. Design Basis Events / Accidents for Engineered Safety Features j (continued) , ! Reactor Service Building (RSB) Not explicitly addressed in Design Filtration System Safety Review Report .
- Steam Generator Building The steam generator buildings are
- Aerosol Release Mitigation designed tn accommodate DBA tempera-System ture transients resulting from sodium spills. The buildings are designed to accommodate DBA i pressure transients resulting from j soditta spills with minimal venting of sodium products to the environment.
Habitability Systems Control room habitability systems I provide for protection from i radiation, toxic chemicals and sodium products of combustion. , Reactor Guard Vessel The LDP incorporates a reactor guard vessel 1 Guard Vessel of PHTS Major The LDP incorporates guard vessels , Components on PHTS major components, the intermediate heat exchangers, and l all piping which is below the l elevation of the tops of the guard vessels. Residual Heat Removal System The LDP incorporates two distinct subsystems: the Steam Generator Auxiliary Cooling System and the Direct Reactor Auxiliary Cooling System (DRACS). The DRACS use natural circulation of the liquid metal for shutdown heat removal. The shutdown heat removal system reliability goal is 3 x 10-3 failures per reactor year. - 78 r ,_ . , _ __ __ _ _ _ , , __,-._-,,.---,.-__y . - -, ..__y%_ ._ ,, _d...., . _ _ _ . _ . . _ . . __ _ _ _ _ _ _ _ _
. TABLE 5. (continued)
CRBR Design Basis Events LDP Desian Basis Events VII. Design Basis Events / Accidents for Engineered Safety Features (continued) Cell Liner Systems The LDP incorporates cell liner
. systems.
Catch Pan and Fire Suppression The LDP incorporates pan and fire Deck System suppression deck systems. e t 79
S. ENRICO FERMI Atomic Power Plant Design basis accidents (DBAs) considered during the design of the
- Enrico FERMI Atomic Power Plant were reviewed against those for the Clinch River Breeder Reactor (CRBR) as one means of ascertaining the completeness of the CRBR DBA delineation. '
~
The review of DBAs for the FERMI plant was based on Section VI, Evaluation of Hazards, of the Technical Information and Hazards Summary Report. Based on this somewhat limited review, the DBA delineation for CRBR appears to be far more' extensive than that for the FERMI plant. One reason may be that the CRBR Preliminary Safety Analysis Report more closely follows the Standard Review Plan (NUREG-0800) which did not exist at the time the FERMI was designed and constructed. Another reason may be that the FERMI plant represented the first scaleup of a fast reactor power plant (it followed ESR-II) and the principal safety concerns were containment of the hypothatical core disruptive accident and fuel and core stability. 4 There were several unique features in the FERMI plant design that are not present in the CRBR design:
- 1. Excess reactivity in the core was limited to preclude ever attaining a super prompt critical condition by rod withdrawal.
- 2. Emergency decay heat removal could be accomplished by reactor heat loss to the below-floor ventilation system.
- 3. A reserve sodium supply was provided in the event leakage into the primary shield tank occurred. This was to prevent interruption of flow in the primary loops. '
4 Water pools were used for storage of irradiated fuel. 80
- 5. A machinery dome was incorporated to restrain the rotating shield plug in the event of a hypothetical core disruptive accident.
Table 6 presents a comparison of the CRBR and FERMI design basis events. O 81 t
i TABLE 6. COMPARISON OF FERM1-1 DBAs WITil Ul0SE FOR CRBR CRBR Design Basis Events FERMI-l Design Basis Events
- 1. Reactivity insertion Events Operational Basis Earthquake Not explicitly discussed in Section VI, Evaluation of llazards,
, Technical Information and llazards Summary Report.
Safe Shutdown Earthquake Withdrawal at Startup Control rod withdrawal speeds and core excess reactivity were limited. Safety rod worth was greater than that for a fuel Withdrawal at power assendily. Small Reactivity Insertions llydrogeneous materials usage was limited (principally for prevention of carburization).
- Ioss of Ilydraulic Iloiddown fuel assendalles held down by mechanical neans (holddown mechanism).
Sudden Core Radial Movement As a result of the EBR-I incident detailed attention was given g to prevention of fuel element bowing and other neans for radial core movenent. Maloperation of Reactor plant plant controller malfunctions were analyzed. It is Controller interesting to note that during the meltdown incident the plant controller automatically compensated for the loss of reactivity which was eventually noticed by an operator. Cold Sodium Insertion The effects of a sudden reduction in coolant inlet temperature were analyzed. Gas Bubble Passage Through Core Other Events Step reactivity insertion equivalent to core excess reactivity was analyzed. The event proceeded slowly enough so that it could be terminated by the safety system. Such a reactivity insertion was judged to be incredible. 4
i
- o e f
i TABLE 6. (Continued) CRBR Design Basis Events FERMI-l Design Basis Events
- 11. Undercooling Events Loss of Offsite Electrical Power The emergency cooling system is capable of removing decay heat even if there were a complete loss of electric power for approximately 24 hours.
4 Spurious Pump Trips Reactor scranwed if more than one primary pump f aiIs. Inadvertent Closure of One Evaporator Not explicitly discussed in Section VI. or Superheater Module Isolation Valve ; Turbine Trip Not explicitly discussed in Section VI. Loss of Normal feedwater An auxiliary feedwater system was provided. Inadvertent Actuation of the Sodium- Addressed in the design analysis. g Water Reaction Pressure Relief System Pump Seizures Addressed in the accident analysis. Small Sodium-to-Water leaks in Steam Provisions for acconnodation of sodium-water leak reactions Generator Tubes taken in design. Sodium-water leaks reactions experienced. failure of the Steam Bypass System Not explicitly addressed in Section VI. ] Steam or Feedline Pipe Break These events were analyzed. Less of Normal Shutdown Cooling Emergency decay heat removal could be accomplished by reactor i System heat loss to the below-floor ventilation system. Large Sodium-Water Reaction Design provisions for large sodium-water reactions. Concepts similar to those for CRBR.
4 T ABL E 6. (Continued) , CRBR Design Basis Events FERMI-l Design Basis Events
) II. Undercooling Events (Continued)
Primary lleat Transport System pipe A full spectrum of PilTS leaks were analyzed. A reserve sodium Leak supply was provided in the event leakage into the primary shield tank occurred to prevent interruption of flow in the primary loops. Intermediate lleat Transport System Intermediate heat transport system piping leak consequences Pipe Leak were analyzed. III. Local failure Events
, Fuel Assendily Particular attention was paid to analyses of flow blockage events. The sudden blockage event (which actually occurred) was also analyzed in detail. The fuel assendily coolant inlet design did not preclude instantaneous blockage.
Control Asscudily Local failure events were analyzed. Radial Blanket Assendily Local failure events were analyzed. IV. Fuel llandling and Storage Events Fuel Assendily Dropped Within Reactor Numerous fuel handling accidents were analyzed and shown to be Vessel During Refueling adequately precluded. Damage of fuel Assendily Due to Not explicitly addressed in Section VI. Attempt to insert a fuel Assendily , Into an Occupied Position . Single Fuel Assendily Cladding f ailure Not explicitly addressed in Section VI. and Subsequent Fission Gas Release During Refueling ; i Cover Gas Release During Refueling Not explicitly addressed in Section VI. i l l l l -
- - . . _ _ - _ = _ _ - - - _.
l . - l TABLE 6. (Continued) CRBR Design Basis Events FERMI-1 Design Basis Events ,r IV. fuel llandling and Storage Events- (Continued) The lleaviest Crane load impacts the Not explicitly addressed in Section VI. Reactor Closure llead Collision of EVIM With Control Rod An EVIM was not used in IERMI. Drive Mechanisms A V. Sodium Spills Primary Sodium In-Containment Storage This is the containment design basis in CRBR. A similar Tank Failure During M.nintenance basis--consumption of all the oxygen in the containment huidling by burning of sodium was used in the FERMI design. Falure of an Ex-Containment Primary Not explicitly addressed in Sectino VI. 4 Sodium Storage Tank l UI Primary lleat Transport System Piping Primary system piping was located in lower half of containment !
. Piping Leaks building which was inerted with nitrogen gas. All primary piping was of double walled. Consequences of major Intermediate lleat Transport System intermediate piping system leaks were used in design analyses.
, Pipe Leak VI. Other Events Loss of DC System Not explicitly addressed in Section VI.
Loss of lustrumentation or Valve Air Not explicitly addressed in Section VI. IllX Leak Intermediate heat exchanger secondary side at higher pressure i than primary side. I j j 1 l j
I ABL E 6. (Continued) CRBR Design Basis Events FERMI-l Design Basis Events VI. Other Events (Continued) Off-Normal Cover Gas Pressure in the Over and underpressure faults were considered in the design and Reactor Coolant Boundary pressure relief and redundant gas supplies and pressure ] regulators were incorporated. Off-Normal Cover Gas Pressure in IHIS Not explicitly addressed in Section VI. Inadvertent Release of 011 Through Pump seal design minimized potential for leakage. Carbonaceous Pump Seal (Pills) and hydrogeneous materials were considered in analysis of reactivity effect and carburization of sodium-wetted metallic surfaces. Inadvertent Release of Oil Through Not explicitly addressed in Section VI. Generator Breaker failure to Open Not explicitly addressed in Section VI. at Turbine frip Rupture in the RAPS Cold Box Not addressed in Section VI. Liquid Radwaste System Failure (Leak Liquid radwaste system fallre consequences were analyzed. I or Rupture) 1 Failure (Leak or Rupture) in the EWST Water pools were used for irradiated fuel storage. NaK Systems l Leakage from Sodium Cold Traps Th.is event was analyzed. 1 Rupture in RAPS Noble Gas Storage Not addressed in Section VI. 1 Rupture in the CAPS Cold Box Not addressed in Section VI. i Leak in a Core Component Pot A similar event was considered in the design of the fuel handling system. y = l
l l ABLE 6. (Continued) CRBR Design Basis Events FERMI-l Design Basis Events i VI. Other Events (Continued) Spent Fuel Shipping Cask Drop From Not explicitly addressed in Section VI. Maximum Possible lleights i ! tiaximum Possible Conventional Fires, These events were considered. Flood, Storms or Miniman River Level , failure of Plug Seals and Annuli These events were considered a machinery dome was added to restain the heat plug in the event of an llCDA. 4 fuel Rod Leakage Combined with IllX Not specifically addressed in Section VI.
; and Steam Generator Leakage i Sodium-Water Reactions in large This event was analyzed.
Conponent Cleaning Vessel Design Basis Events / Accidents for Engineered Safety features Reactor Confinement / Containment The FERMI-l containment was designed to withstand the effects of an energy release equivalent to 1000 lb of TNT at the core location. A machinery dome was used to prevent the top shield plug from penetrating the containment. Containment isolation System Containment location systems were provided. Annulus Filtration System N/A for FERMI-1. Reactor Service Building (RSB) N/A for FERMI-1. Filtration System Steam Generator Building Aerosol N/A for FERMI-1. l Release Mitigation System liabitability Systems Not explicitly addressed in Section VI. j
TABLE 6. (Continued) CRBR Design Basis Events FERMI-l Design Basis Events VI. Other Events (Continued) Design Basis Events / Accidents for Engineered Safety Features (Continued) Reactor Guard Vessel A reactor guard vessel per se was not used. The primary shield tank enclosed the reactor vessel but the volume of Guard Vessels of PilTS Major sodium that would be lost from the reactor vessel in the event Components of a leak required a reserve sodium supply to preclude interruption of flow in the primary loops. Residual lleat Removal System Only one loop out of three required for shutdown heat removal. If feedwdter lost reactor heat losses to the below floor ventilation system would provide emergency cooling. Cell Lin r System Liners were employed to provide leak tightness and prevent sodium-concrete reactions. co Catch Pan and Fire Suppression Not employed in the FERMI plant. Deck System Other Events Analyzed A hypothetical meltdown accident equivalent to energy release from the detonation of 1,000 lbs of TNT was the basis for designing the reactor containment. s
- w %
IV. CONCLUSIONS l i To the extent that a somewhat limited review of design basis accidents for U.S. fast reactors can provide confidence in the completeness of the delineation of design basis accidents for CRBR, it appears that the breadth and depth of CRBR design basis events is greater than that for the early plants (EBR-II, SEFOR, FERMI) and closely approximates that for FFTF and
^
LDP. This is as one would expect if the collective experience and judgement is being properly employed in design basis accident delineation. l A continuing study of design basis accident delineation for CRBR has been incorporated in the applicants probabilistic risk assessment 'or CRBR. This work, which is in progress as part of the applicants voluntary compliance with the pertiner.t portion of NUREG-0718, will be completed prior to the issuance of an operating license. Should the study reveal event sequences which have not received sufficient attention during the design of the plant, they will be addressed prior to issuance of the operating license. 89
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