ML20035G677
| ML20035G677 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 04/23/1993 |
| From: | Poslusny C Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9304290053 | |
| Download: ML20035G677 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION h
WASHINGTON, D. C. 20S55
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April 23,1993 Docket No.52-001 APPLICANT:
PROJECT:
Advanced Boiling Water Reactor (ABWR)
SUBJECT:
SUMMARY
OF MEETING 0F APRIL 15, 1993 A public meeting was held between GE and Reactor Systems Branch staff on April 15, 1993, to discuss design certification aspects of the fuel design for the ABWR. is a list of those who attended the meeting.
GE provided a draft copy of the Tier 1 design descriptions for the nuclear fuel and the control blade designs (Enclosure 2). The staff provided comments on the fuel bundle description, Section 2.8.1, as follows. Regarding the design description, it was suggested that the first paragraph be deleted with the exception of the last sentence, and that the phrase " utilizing the certified design" also be deleted.
For the thermal-mechanical and nuclear sections, it was suggested that the phrase "NRC approved methods and criteria" be deleted and the criteria descriptions be revised to be more consistent with standard review plan guidance.
GE staff stated that the inspections, tests, analyses, and acceptance criteria (ITAAC) section for the fuel design had been deleted from the Tier 1 informa-tion based on GE's understanding of the staff requirements. The staff indicated in the meeting that its position is that an ITAAC is needed for the reactor fuel to verify the required analyses for fuel performance has been completed by the licensee prior to 'uel load.
Given this requirement in the ITAAC, the staff stated that the specific methodology which GE used to evaluate the fuel, and the results of the analyses for the fuel design, should be included in the SSAR.
GE stated that this information had not been provided in the standard safety analysis report (SSAR) because it had not been called out as a specific requirement in the staff's draft final safety evaluation report as an open or confirmatory item.
The staff indicated that its position was that GE needed to provide in its application a " standard Chapter 4" description of the fuel design which includes a detailed description of the evaluation methods and the detailed analysis results for the ABWR fuel.
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. April 23, 1993 It was agreed that GE would consider adding the identified information to the ITAAC and to the SSAR, and the staff would reconfirm its positions with the Office of Nuclear Reactor Regulation management. A subsequent phone call would be conducted to discuss these items further.
(Original signed by)
Chester Poslusny, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosure:
See next page DISTRIBUTION w/ enclosures:
Docket File PDST R/F DCrutchfield PDR CPoslusny SNinh PShea DISTRIBUTION w/o enclosures:
TMurley/FMiraglia WRussell JMoore, 15B18 GGrant, 17G21 EJordan, MNBB3701 J0'Brien, RES LShao, RES RBorchardt s
LPhillips, 8E23 RJones, 8E23 HRichings, 8E23 GThomas, 8E23 MFinkelstein,0GC TBoyce
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LA:PDST:ADAR PM:PDST:ADAR SRXB SC:PDST:ADAR Ogg93'Q)
NAME: PShea O CPoslusny:tz LPhillips JNWilson DATE:
04/ /93 04/g3/93 04/13/93 0FFICIAL RECORD COPY: MS415.CP a
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. April 23, 1993 It was agreed that GE would consider adding the identified information to the ITAAC and to the SSAR, and the staff would reconfirm its positions with the Office of Nuclear Reactor Regulation management. A subsequent phone call would be conducted to discuss these items further.
a h
Chester Poslusny, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
As stated i
cc w/ enclosure:
See next page l
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GE Nuclear Energy Docket No.52-001 cc:
Mr. Patrick W. Harriott, Manager Mr. Joseph Quirk Licensing & Consulting Services GE Nuclear Energy GE Nuclear Energy General Electric Company 175 Curtner Avenue 175 Curtner Avenue, Mail Code 782 San Jose, California 95125 San Jose, California 95125 Mr. Robert Mitchell General Electric Company 175 Curtner Avenue l
San Jose, California 95125 Mr. L. Gifford, Program Manager i
Regulatory Programs GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs U. S. Environmental Protection Agency 401 M Street, S.W.
Washington, D.C.
20460 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.
20585 l
Mr. Steve Goldberg i
Budget Examiner i
725 17th Street, N.W.
Room 8002 Washington, D.C.
20503 i
Mr. Frank A. Ross
^
U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 j
Mr. Raymond Ng 1776 Eye Street, N.W.
Suite 300 Washington, D.C.
20006 Marcus A. Rowden, Esq.
Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.
Suite 800 Washington, D.C.
20004 j
Jay M. Gutierrez, Esq.
Newman & Holtzinger, P.C.
1615 L Street, N.W.
Suite 1000 Washington, D.C.
20036 1
P ABWR FUEL MEETING t
ATTENDEES APRIL 15, 1993 NAME AFFILIATION Chet Poslusny NRR/PDST Howard Richings NRR/SRXB Michael Finkelstein 0GC/ Hearings / Reactors George Thomas NRR/SRXB Robert C. Jones NRR/SRXB Tom Boyce NRR/PDST Larry Phillips NRR/SRXB Jim Klapproth GE/ Fuel Licensing 7
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ABWR Design Document h2 oft ~
2.8 Nuclear Fuel 2.8.1 Fuel' Bundle Design Description Fuel design for the ABWR is not within the scope of the certified design. It is intended that the specific fuel to be utilized in any facilitywhich has adopted the certified design be in compliance with U.S. NRC approved fuel design criteria.
This strategy is intended to permit future use of enhanced / improved fuel designs as they become available. However, this approach is predicated on the assumption that future fuel designs will be extensions of the basic fuel technology that has been developed for boiling light water reactors. Key characteristics of this established BWR fuel technology are:
(1) Uranium oxide-based fuel pellets.
(2) Zirconium-based (or equivalent) fuel cladding.
(3) Material selected on the basis of BWR operating conditions.
(4) Multi-rod fuel bundles.
(5) Fuel bundle inlet orificing to control bundle flow rates, core flow distribution, and reactor coolant hydraulic characteristics.
The following is a summary of the principal requirements which must be met by the fuel supplied to any facility utilizing the certified design.
Thermal Mechanical The fuel bundle is evaluated using NRC approved methods and criteria to assure that:
(1) Fuel rod failure does not occur as a result of normal operation and anticipated operational occurrences.
(2) Fuel bundle damage is never so severe as to prevent control rod insertion when required.
(3) The number of fuel rod failures is not underestimated for postulated accidents.
(4) Coolability is always maintained.
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bhW ABWR Design Document
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Nuclear The fuel bundle is evaluated using NRC approved methods and criteria to assure that:
(1) A negative Doppler reactivity coefficient is maintained.
(2) A negative core moderator void reactivity coefficient resulting from boiling in the active flow channels is maintained for design basis operating conditions.
(3) A negative moderator temperature coefficient is maintained above hot standby.
(4) For a super prompt critical reactisityinsertion accident, the net prompt reactivity feedback due to prompt heating of the moderator and fuelis negative.
(5) A negative power coefficient, as determined by calculating the reactivity change due to an incremental power change from a steady-state base power level, is maintained for operating power levels above hot standby.
(6) The core is capable of being made subcritical in the most reactive condidon throughout an operating cvcle with the most reactive control rod in its fu4out position and all other rods fully inserted.
Hydraulic The reactor core and associated coolant, control and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
Stability compliance considers potential limit cycle response within the limits of safety system and/or operator intervention, and assures that specified acceptable fuel design limits are not exceeded.
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%W ABWR oesign Docum:nt r~
.We 2.8.2 Fuel Channel Design Description Fuel channel design for the ABWR is not within the scope of the cerdfied design.
The key characteristic of this established BWR fuel channel technologyis the use of zirconium-based (or equivalent) fuel channels which preclude cross-flow in the core region.
The following is a summary of the principal requirements which must be met by the fuel supplied to any facility utilizing the certified design.
The fuel channel design is evaluated using NRC approved methods and criteria to assure that:
(1) Fuel channel damage is never so severe as to prevent control rod insertion when it is required.
(2) Coolability is always maintained.
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-1 2.8 2
ABWR oe:ign Docum:nt
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2.8.3 Control Blade Design Description Control blade design for the ABWR is not within the scope of the certified design. Key characteristics of this established BWR control blade technology are:
(1) Control blades perform dual functions of power distribution shaping and reactisity control.
(2) The control blade has a cruciform cross-sectional envelope shape.
(3) The control blade has a coupling at the bottom for attachment to the control blade drive.
- 4) The control blade contains neutron absorbing materials.
The following is a summary of the principal requirements which must be met by the control blade supplied to any facility utilizing the certified design.
(1) The control blade stresses, strains, and cumulative fatigue are evaluated to not exceed the ultimate stress or strain of the material.
(2) The control blade is evaluated to be capable ofinsertion into the core within the limits specified in Section 2.2.2.
(3) The material of the control blade is compatible with the reactor environment.
l (4) The reactivity worth of the control blade shall be included in the plant core analyses.
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