ML20246M413
| ML20246M413 | |
| Person / Time | |
|---|---|
| Issue date: | 05/10/1989 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| Shared Package | |
| ML20246F804 | List: |
| References | |
| TASK-PII, TASK-SE SECY-89-153, NUDOCS 8905190134 | |
| Download: ML20246M413 (7) | |
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May 10, 1989 POLICY ISSUE SECY-89-153
_For: The Commissioners O N M.
From: Victor Stello, Jr.
Executive Director for Operations
Subject:
SEVERE ACCIDENT DESIGN FEATURES OF THE ADVANCED BOILING WATER REACTOR (ABWR)
Purpose:
In the statement of considerations of 10 CFR Part 52, the Commission asked the staff to advise them on the need for criteria that are different from or supplementary to current standards. The purpose of this paper is to inform the Commission of certain features of General Electric's (GE's) ABWR design that the staff believes will enhance safety and will satisfac-torily 3ddress severe accident issues when the staff's review is complete.
Background:
SECY-89-013 informed the Commission of recent staff decisions rsgarding design enhancements for three standardized evolutionary advanced light-water reactors (ALWRs) -- General Electric's ABWR, Westinghouse's RESAR SP/90, and Combustion Engineering's CESSAR System 80+ -- and the Electrir. Power Research Institute's (EPRI's) Utility Requirements Document.
Consistent with the Commission policy on severe 6ccidents, General Electric has proposed additional design featu'es to enhance the capability of the ABWR to prevent and mitigate severe accidents. These enhancements were developed by GE in the context of their commitments in the Advanced Boiling Water Reactor Licensing Review Bases ( ABWR-LRB), dateo August 7,1987, to further reduce the core damage frequency as well as the potential for a large release for this design.
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Contact:
C. Miller, NRR/FDSNP 492-1118
r The Commissioners ,
y Discussion: The severe accident issues to be dealt with during the reviews of the ABWR design are briefly discussed below. A certain
-number of these issues were previously discussed in SECY-89-013.
The severe accident concerns as they relate:to the other designs,
-Westinghouse's RESAR SP/90, and Combustion Engineering's System 80+,
will be discussed with the Commission at a later date.
In the.ABWR design, GE has committed to the following Commission guidance on severe accidents for future plants as codified by 10 CFR Part 52:
- Demonstration of compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f);
- Proposed technical resolution of the applicable Unresolved Safety Issues and medium- and high-priority Generic Issues; and.
- A design-specific probabilistic risk assessment.
This pap' .s structured in such a marner to identify (1) certain major initiating events that could lead to a severe reactor accident,(2)ABWRdesionenhancementsthatreducetheprobability oftheiroccurrence,(3)importantABWRsystemsthatwouldbeused tomitigatea-severeaccidentshouldoneoccur,and.(4)other topics reltted to severe accidents and the ABWR design. The initiating events and severe accident issues identified have resulted from a thorough staff review of past severe accident studies and research applicable to the ABWR.
STATIO!! BLACK 0UT The statinn blackout rule (10 CFR 50.63) allows utilitics several design alternatives to ensure that an operating plant can safely shut down in the event that all ac power (offsite andonsite)islost.
The staff believes that the preferred method of demonstrating compliance with 10 CFR 50.63 is through the installation of an
) alternate ac power source installed spare (100 percent capacity (from a different manufac-(andauxiliarie:)ofdiversedesign turer) that is consistent with the guidance in Regulatory Guide 1.155.
The ACWR design clearly goes beyond the requirements identified in the station blackout rule. The ABWR design includes three independent electrical divisions, each with high-pressure and low-pressure capability, each powered by a 100 percent capacity diesel generator, and each division capable of independently shutting the reactor down. Additionally, the ABWR design includes an alternate ac combustion turbine to back up the
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l diesel generators. The design has a capability to survive a ;
10-hour blackout perioo utilizing the reactor core isolation ;
cooling (RCIC) turbine and station batteries. Extended blackout l capabilities-are also provided by the ac-independent water i addition system. This system allows for makeup to the reactor vessel following'RCS depressurization by connecting a direct ;
drive diesel fire pump or by connecting an external pumping 1 source, such as a fire truck, to a yard standpipe.
INTERSYSTEM LOCA j Future 'ALWR designs can reduce the possibility of a' loss-of-coolant accident (LOCA) outside containment by designing (to the ;
extent practicable) all systems and subsystems connected to the i reactor coolant system (RCS) to an ultimate rupture strength at l 1 east equal to the full RCS pressure. 1 1
The ADWR has been designed to minimize the possibility of an interfacing system LOCA in the following ways. The low pressure- ,
systems directly interfacing with the RCS are designed with 500 l psig piping which provides for a rupture pressure of approximately 1000 psig. In addition, the high/ low-pressure motor-operated isolation valves have safety-gr6de, redundant pressure interlocks. !
Also, the motor-operated emergency core cooling system (ECCS) valves will only be tested when the reactor is at low pressure.
All inboard check valves on the ECCS will be testable and have 1 position indication. Additionally, design criteria used by GE l require that all pipe designed to 1/3 or greater of reactor ;
pressure requires two malfunctions to occur before the pipe ;
would be subjectea to reactor system pressure. The pipe der ~igned ,
to less than 1/S reactor pressure requires at least three malfunctions before the pipe would be subjected to reactor system 1
pressure.
The staff believes that these features should provide sufficient ;
margin for all high/ low-pressure interfaces to eliminate the concern about LOCAs outside of containment at the high/ low-pressure interface of systems connected to the reactor coolant system (RCS).
l ANTICIPATED TRANSIENT WITHOUT SCRAM i The anticipated transient without scram ( ATWS) rule (10 CFR 50.62) was promulgated to reduce the probability of an ATWS event and to enhance mitigation capability if such an event occurred. In this regard, the ABWR design has a number of features that reduce the risk from an ATWS event. The staff believes that the modest enhancements proposed by GE can further i' reduce the risk from an ATWS event. These features include a diverse scram system with both hydraulic and electric run-in t
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capability on the control rods, a manuel-operated standby liquid control system, and a recirculation pump trip. capability. In-addition, the scram discharge volume has been removed from the AEWR,_ eliminating some of the potential ATWS problems associated with the older BWR designs.
uires an automatically initiated standby While liquid the ATWS control rule req (SLCS), GE has concluded that the enhanced system reliability of the reactor protection system negates the need for an automatic SLCS. General Electric has agreed to provide a reliability analysis of the SLCS to support this position. The staff believes the anelysis should include an evaluation of the control room displays, emergency procedures, and the time available for operator action.
HYDROGEN CONTROL The Commissicn's Severe Accident and Standardization Policy Statements stated that future designs should address the provisions of 10 CFR 50.34(f). The Commission's stated policy has'been codified in 10 CFR Part 52 to require the technically
. relevant prveisions be met. Specifically, in order that contain-mentintegritybemaintained,-10CFR50.34(f)(2)(ix) requires future Cesigns to prov;de.a system for hydrogen control'that can safely a:commodate hydrogen generated by the equivalent of a 100 percent fuel-clad metal water reaction. In addition, this system needs to De capable of precluding uniform concentrations of hydrogen ihm exceeding 10 r,ercent (by volume), or an inerted atmosphere witnie, the containment must be provided.
The APWR design meets the requirements of 10 CFR 50.34(f)(2)(ix) by utilizing a nitropn-5nerted atmosphere within containment.
Also, a hydrogen recombiner for design-basis accidents will be provided in the ABWR design.
l VENTING In order to preclude an irreversible rupture of the containment, the ABWR design will include a " hardened" wetwell vent capable of venting at pressures up to approximately 80 psig. Venting operations will require the use of de power and pneumatic pressure to open the isolation valves. In venting situations, all fission products would be filtered through the suppression ;
pool. The containment vent system will prevent containment failure due to overpressure and temperature.
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%1 -The Commissioners' ' 5-CORE-CONCRETE INTERACTION - ABILITY,TO COOL CORE DEBRIS i In the unlikely event of a severe acculent in which the core L has melted through the reactor vessel, it is possible that l containment integrity could be breached if the molten core is not sufficiently cooled. In addition, interactions between the core debris and concrete-can generate large quantities of additional y hydrogen and other non-condensible gases.
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! The ABWR design has a number of' features that would mitigate the effects of a molten core. The suppression pool surrounds the lower drywell. cavity and would thereby prevent core debris ~
from reaching the containment boundary and breaching its integ-rity. Also, the ABWR is designed with a lower drywell flooder and a ~ cavity space sufficientp to be able .to disperse core debris at an energy level of 1 MWt/m . The flooder consists of a number of temperature-sensitive. fusible plugs that allow suppression pool water to enter the drywell cavity when high temperature resulting from core debris occurs in the lower drywell. The horizontal vents to the suppression pool will remain covered in the event of lower.drywell flooding. GE anticipates that any core-concrete interaction will be stopped when the suppression pool water quenches the molten core debris. . By providing sufficient area to allow the core debris to spread to a shallow bed and by flooding the core debris, it is expected that the potential for extensive core-concrete interactions will be significantly reduced. In addition, even if limited core-concrete interactions continue, the overlying poo/ of water will mitigate the consequences of these interactions ty scrubbing the fission products and cooling the gases released 1 rom the concrete.
SOURCE TERMS Regulatory Guides 1.3 and 1.5 provide the staff's principal bases for implementing the requirements in 10 CFR Part 100 for the ABWR. The radiological " source term" in these guides is based, in part, on the 1962 " TID-14844 source term." From the outset, these in-containment source terms were widely acknowledged to be very conservative but were justified on the basis of the uncertainties associated with accident sequences and equipment performance at the time of promulgation (circa 1962).
The staff intends to continue using existing methodology for the licensing basis source term for ALKRs in order to confirm ;
compliance with 10 CFR Part 100 with respect to design basis accidents. The ABWR design will provide dual remote control room ventilation intakes and will maintain a combined main steam isolation valve (MSIV) leakage rate of .less than 100-150 standard cubic feet per hour (scfh). The staff will consider giving credit for the steam line condenser as a hold-up pathway, and to the i
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[ acceptance of a single charcoal bed filter in the standby gas.
treatment system. This. credit, however, is predicated on additional analysis by GE regarding these systems and subsequent staff review.
The staff is presently working with GE in order to agree on a -
design specific source term and a definition of containment L failure that can be used to address the severe accident goals
' identified by GE. These goals limit the probability of occur-renceofoffsitedosesinexcessof25rgmbeyondaone-halfmile and a containm nt radius from failure conditional. the reactor to less probability than less 10~
than 10/yeg, weighted over credible core damage sequences.
g PROBABILISTIC RISK ASSESSMENT (PRA) 10 CfR Part 52 requires a design-specific PRA based upon the bounding. site parameters for the design. To ensure the design certification PRA assumptions are retained as part of the design and operation of the plant, the ABWR licensing bases will be incorporated into the design certification.
GE has committed to provide a level-3 PRA including full power, low power and refueling conditions and addressing internal events and a. bounding external events analysis. GE has also agreed to provide in the ABWR standard safety analysis report (SSAR) the reliability and maintenance criteria that a future applicant must satisfy to ensure that the safety of the as-built facility will continue to be accurately described by the certified design. The ABWR SSAR is to include the key assumptions of the PRA and other PRA licensing commitments.
BWR THERMAL-HYDRAULIC STABILITY The staff believes the issue of BWR stability, while not directly a sevure accident concern for the ABWR, is an important topic relative to design certification and is therefore discussed in this paper.
In boiling-water reactors, thermal-hydraulic instabilities can cause oscillations that can result in violation of the minimum critical power ratio (MCPR) safety limits. In order to cope with the stability problem, GE has provided additional preventive and mitigative measures to the ABWR design. These
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- 2 lw ; e The Commissioners 4-include, in part, an automatic logic that prevents plant operation in the region of least stability, and selected control rod run-in that is automatically initiated to avoid stability concerns when a trip of two or more reactor internal pumps (RIPS) occurs. The ABWR design reduces the potential for the type of event that occurred at LaSalle.
Conclusion:
The staff believes that the issues and associated ABWR design enhancements discussed in this paper reflect the experience that has been gained from the current generation of operating plants, ni.d that they are in keeping with the Commission's policy that isture designs for nuclear pler.ts should reduce the risk from severe accidents. The staff belieees that its review will confirm the effectiveness of these features in addressing the Commission's severe accident goals defined in the Severe Accident, Safety Goal, and Standardization Policy Statements as well as complying with 10 CFR Part 52 as it relates to design certification.
It is the staff's intention that equipment required to mitigate a severe accident but not required to mitigate a design-basis accident (DBA) (e.g., core cebris quenching equipment) must be able to maintain its intended function for as long as required but need.not be safety grade or single-failure proof.
Further, the staff intends to address compliance with the severe accident requirements defined in Commission policy and 10 CFR Part 52 on a design-specific basis rather than through generic rulemaking as' described in SECY-88-248. The staff believes that this approach will minimize scheduling impacts on the individual designs that will ultimately culminate in specific rules through the design certification process.
The staff does not envision any additional requirements to be imposed on the ABWR at this time; however, it will ensure that GE design commitments are sufficient through the detailed review of the design. The staff will inform the Comission during the design certification process if additional requirements are determined necessary for the ABWR design to comply with the Comission's severe accident requirements.
A copy of this paper has been provided to the ACRS so that they might provide their views to the Commission. The staff is scheduled to brief the ABWR Subcommittee on this subject on flay 10-11, 1989.
. DISTRIBUTION:
Commissioners EDO .- j
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OGC IG ACRS 44 j' f ACNW GPA ASLBP Victor Stello J r. I Executive Director REGIONAL OFFICES ASLAP for Operations SECY
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- . UNITED STATES IN RESPONSE, PLEASE.
!" n NUCLEAR REGULATORY COMMISSION REFER TO: M890620 j-l' -( j W ASHINGT ON, D.C. 20555 g j$ July 31, 1989
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OFFICE OF THE
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l MEMORANDUM FOR: James M. Taylor Acting Executive Director for Operations Raymond F. Fraley, Executive Director Advisory Committee on R tor Safeguards FROM: Samuel J. Chilk, Secret [
SUBJECT:
STAFF REQUIREMENTS - BRIEF:ING ON THE APPLICATION OF THE SEVEhE' ACCIDENT POLICY TO THE LEAD APPLICATION FOR ADVANCED LIGHT WATER REACTORS, 10:00 A.M., TUESDAY, JUNE 20, 1989, COMMISSIONERS' CONFERENCE ROOM, ONE WHITE FLINT NORTH, ROCKVILLE, MARYLAND (OPEN TO PUBLIC ATTENDANCE)
The Commission.was briefed by the staff on the review status of General Electric's (GE) submittal of an Advanced Boiling Water Reactor (ABWR) for design certification. The briefing included a discussion of proposed methods for resolving severe accident and other safety-issues.
In order to assist the commission in making policy decisions, the Commission' requests the staff to:
- a. Submit to the Commission a paper describing the status of efforts to develop an updated source term analytical methodology that takes into account current knowledge on the subject. Discuss the extent to which the current deterministically-established source term (TID-14844) can be updated or otherwise improved, based'upon the knowledge now available, while still adhering to the deterministic approach. Address constraints of any kind which preclude ,
regulatory application of an updated, more realistic source term in the licensing basis for future reactors, 1 I
including the GE ABWR, such as implications for other areas of the Commission's regulations currently based upon or affected by the TID-14844 source term. Explain how the )
schedule for any update of the source term is tied to current containment performance studies. Where the staff is of the view that uncertainties in current knowledge exist, the staff should discuss these uncertainties and explain the significance of any such uncertainties with c>a rr o s> A n e t J 3 }-_V # !_N 5
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lsi lsi respcct.to' potential regulatory applications'of the
-updated source term 1 information. If the existing source term (TID-14844)Lis-used for the.ABWR-licensing basis, discuss the need for any departure from the approaches set y .forth in the Standard Review Plan or relevant Regulatory t Guides-for the calculation of offsitefdoses in licensing basis analyses (e.g.,~giving credit for non-safety related -)
l; equipment for fission' product retention).
' (EDD) (SECY Suspense: 10/30/89)~
- b. Submit to the Commission a paper describing:
l l 1. . Developments since September 6, 1988, which have' led the staff to conclude that establishing severe
- accident requirements for future reactors, including the ABWR, by. generic rulemaking (as described in SECY-88-248 and staff's December 1, 1988. memorandum)
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.is no longer the preferred approach. Include a summary of the two workshops.that have been conducted on this subject. Describe how these developments have'affected or otherwise altered'each of-the policy, technical, legal, and schedular consider-ations discussed in SECY-88-248.- Address the impact of formally preparing a rule in parallel with the standard plant reviews. Include the updated schedules for review of the GE'ABWR, the Combustion Engineering System 80+, and the Westinghouse SP/90.
- 2. The severe accident issues, based upon current knowledge including the-staff's review of'the-GE ABWR to date, that staff is proposing to be addressed in the applications for future reactors. Include the criteria staff proposes to'use to judge.the accept-ability of a future design with respect to each issue.
- 3. The measures to ensure that systems and equipment
. required only to mitigate severe accidents are available to perform their intended function (e.g.,
environmental qualification, etc.).
(EDO) (SECY Suspense: 9/15/89)
- c. . Submit to the Commission a paper describing:
- 1. Those instances to date where, in the review of the GE ABWR and in'the discussions with GE, the staff l would propose to go beyond what is currently required by the regulations or the Standard Review Plan, as well as those instances where the staff uould propose an approach that does not go as far as either current
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regulations or the Standard Review Plan would require (e.g., the relationship of the operating basis earthquake to the safe shutdown earthquake, station j blackout requirements, etc.). In each such instance, the staff should explain whether the proposed approach involves an issue that is unique to the GE l ABWR or that is generic in nature, and the basis for j the staff's conclusion. j
- 2. The proposed ABWR containment vent design. Would the !
ABWR meet the Commission's Safety Goal and the proposed ABWR severe accident goals without such a j system? Had this design not been proposed, what is <
staff's thinking on the alternative means for assuring containment integrity per 10 CFR 50.34(f)? Describe the pros and cons to staff's and applicant's proposals /
options. Explain the pros and cons in terms of generic applicability for all future reactors. The paper should also describe the basis for, and value of, the proposed conditionaly containment failure probability criterion of 10
- 3. The status and schedule for GE's submittal of its reliability and maintenance criteria for the ABWR.
Address the standards the staff intends to use in the review of such criteria. Provide thoughts and recommendations, beyond those documented in NUREG 1333, on options for improving maintenance and reliability for the future reactors, taking into consideration U.S. and foreign experiences, along the lines of the Japanese maintenance outage programs.
(EDO) (SECY Suspense: 10/30/89)
- d. The Commission requests a copy of the safety evaluation report on the GE reliability and maintenance program prior to issuance.
(EDO) (SECY Suspense: When ready)
- e. Keep the ACRS informed of ongoing activities in order to assure timely ACRS comments to the Commission.
(EDO) (SECY Suspense: As required)
The Commission requests the ACRS to submit a status report to the Commission which:
- 1. Describes the scope and schedule for the ACRS effort to develcp criteria for containment designs for future reactors; and
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- 2. Compares the criteria under consideration by the ACRS with l those that the staff would propose to apply, identifying i
any differences or inconsistencies.
l (ACRS) (SECY Suspense: 10/13/89)
The staff outlined the following tentative schedule for its review of the GE ABWR certification application:
- 1. Issue in the Spring of 1990, a final draft safety evaluation report for review and comment by the Advisory Committee on Reactor Safeguards.
- 2. Issue in the Summer of 1990, a final safety evaluation report and a final design acceptance report.
- 3. Initiate in late Summer 1990, a hearing on the certification submittal.
- 4. Issue in October 1991, the design certification for the ABWR.
An updated schedule which incorporates the tasks outlined in this SRM will be provided in the Commission paper due 9/15/89 (item "b" above).
(EDO) (SECY Suspense: 09/15/89) cc: Chairman Carr !
Commissioner Roberts Commissioner Rogers ;
Commissioner Curtiss l OGC GPA PDR - Advance DCS - Pl-24 t
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