ML20046C390
| ML20046C390 | |
| Person / Time | |
|---|---|
| Issue date: | 06/20/1993 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Rasin W NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| References | |
| NUDOCS 9308100241 | |
| Download: ML20046C390 (12) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 i'.....f June 20,1993 Mr. William H. Rasin Vice President and Director Technical Division Nuclear Management and Resources Council 1776 Eye Street, N.W., Suite 300 Washington, D.C.
20006-3706
Dear Mr. Rasin:
SUBJECT:
DESIGN CERTIFICATION ISSUES I am responding to your March 12, 1993, letter in which you commented on the staff's views of "roadmaps" and the role of the design control document (DCD) in a design certification rule.
I recognize that the Nuclear Management and Resources Council (NUMARC) is committed to providing industry input to deci-sions involving the implementation of 10 CFR Part 52, and the staff has discussed these important issues with you.
The vendors are submitting to the staff various cross references, or "roadmaps," between the standard safety analysis report (SSAR) and the Tier I design certification material. These roadmaps describe how the vendors have incorporated key features of the design into the inspections, tests, analyses, and acceptance criteria (ITAAC).
You expressed a concern that inclusion of these roadmaps in the DCD could potentially create a procedural relationship between Tier 2 and the Tier I design information.
You further surmised that the roadmaps could potentially create the implication that a significant portion of the material in Tier 2 was subject to the stringent change procedures for Tier 1 design material.
The staff believes that roadmaps do not create any more procedural relation-ships than are already called for by Part 52 in the "50.59-like" change process. As discussed in SECY-92-287, " Form and Content for a Design Certifi-cation Rule," the process to change a facility specified in the "50.59-like" process permits a combined license (COL) applicant or licensee to make a change to a facility unless the proposed chance involves a chance to the certified standard design, the technical specifications, or an unreviewed safety question.
The cross references merely provide a clear "roadmap" for the staff, COL applicant, or licensee showing what design features in the certified standard design and supporting analyses in the SSAR must be considered as part of the 50.59 evaluation, prior to making any changes to a facility.
The roadmaps in no way inhibit the change process for Tier I and r
Tier 2 design information.
g /, M7 Part 52 requires the ITAAC to provide reasonable assurance that a facility d4[#
is built and will operate in accordance with the approved design and applicable regulations.
SECY-92-287 discussed the submittal of roadmaps as y p F 7 part of the SSAR.
Further, as stated in my letter to you of November 25, pe#6 1992, roadmaps should be retained in the Tier 2 design information in the SSAR heduc ) '
analyses for a design are adequately verified by the ITAAC, and to ensure the(
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l and the DCD to provide reasonable assurance that the design basis and key "50.59-like" change process adequately considers the certified standard design during the lifetime of a facility referencing the standardized design.
There p J 47 l
9308100241 930620
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Mr. William H. Rasin June 20, 1993 has been no new information developed since that letter that would cause the staff to reevaluate its views or where the roadmaps should ultimately be retained.
In addition to the reasons cited above, the roadmaps have other significant, practical uses.
These are discussed in Enclosure 1.
A discussion of the format and content of these roadmaps is contained in Enclosure 2.
A list of the current list of analyses to be roadmapped is contained in Enclosure 3.
You zlso expressed your views that the DCD should be decoupled from the issuance of a final design approval for the designs.
The staff's position is provided in SECY-93-097, " Integrated Review Schedules for the Evolutionary and Advanced Light Water Reactor Projects," of April 14, 1993.
The staff is also preparing a paper discussing the form and content of the DCD.
We will continue to interact with industry representatives in the development of Part 52 implementation issues. We share the common goal of expeditious development of these concepts in order to successfully implement the goals of 10 CFR Part 52.
Sincerely, (Original signed by Robert C. Pierson for)
Dennis M. Crutchfield, Associate Director for Advanced Reactors and License Renewal Office of fluclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures:
See next page DISTRIBUTIO!!:
Central File PDST R/F TMurley DCrutchfield PDR WRussell Jf4Wil son RBorchardt JMoore, 15B18 PShea FHasselberg TKenyon ACRS (11) (w/o encl. )
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MMalloy Stiinh CPoslusny TEssig AThadani, 8E2 RBarrett, BH7 WBeckner, 10E4 RJones, 8E23 DTiratcher, 7E4 DTerao, 7E2 CMcCracken, 8D1 MRubin, 8E2 g g,D b 3 0FC:
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ROADMAPS.TB
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Mr. Robert Mitchell Mr. Joseph Quirk General Electric Company GE Nuclear-Energy 175 Curtner Avenue General Electric Company San Jose, California 95125 175 Curtner Avenue, Mail Code 782 San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 s
Rockville, Maryland 20852 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.
20585 Mr. Steve Goldberg l
Budget Examiner 725 17th Street, N.W.
i Room 8002 Washington, D.C.
20503 Mr. B. A. McIntyre Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit P.O. Box 355 Pittsburgh, Pennsylvania 15230 l
Mr. John C. Butler Advanced Plant Safety & Licensing Westinghouse Electric Corporation i
Energy. Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 1
Mr. M. D. Beaumont t
Nuclear and Advanced Technology Division i
Westinghouse Electric Corporation l
One Montrose Metro 11921 Rockville Pike Suite 350 Rockville, Maryland 20852 Mr. S. M. Modro EG&G Idaho Inc.
Post Office Box 1625 Idaho Falls, Idaho 83415 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 l
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cc:
Director, Criteria & Standards Division Office of Radiation Programs U. S. Environmental Protection Agency 401 !! Street, S.W.
Wash 1ogton, D.C.
20460 fiarcus A. Rowden, Esq.
Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.
Suite 800
'!::.W;t:n, D.C.
20004 Jay M. Gutierrez, Esq.
Newman & Holtzinger, P.C.
1615 L Street, N.W.
Suite 1000 Washington, D.C.
20036 Mr. C. B. Brinkman, Acting Director Nuclear Systems Licensing ABB-Combustion Engineering, Inc.
1000 Prospect Hill Road Windsor,' Connecticut 06095-0500 Mr. C. B. Brinkman, Manager Washington Nuclear Operations ABB-Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch Nuclear Systems Licensing ABB-Combustion Engineering, Inc.
1000 Prospect Hill Road
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Post Office Box 500-Windsor, Connecticut 06095-0500 Mr. Raymond Ng 1776 Eye Street, N.W.
Suite 300 Washington, D.C.
20006 Joseph R. Egan, Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.
Washington, D.C.
20037-1128 Mr. Regis A. Matzie, Vice President Nuclear Systems Development ABB-Combustion Engineering, Inc.
1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500
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l ADDITIONAL CONSIDERATIONS 1.
As stated in SECY-92-287, " Form and Content for a Design Certification Rule," roadmaps ensure that the key analyses and features of the design on which the staff's safety findings are based are not eroded during "50.59-like" change process. This answers the Commission's SRMs on SECY-90-377,
" Requirements for Design Certification Under 10 CFR Part 52," (Level of Detail), SECY-91-178, " Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications and Combined Licenses,"
(Concept of ITAAC), and SECY-92-053, "Use of Design Acceptance Criteria During 10 CFR Part 52 Design Certification Reviews," (Concept of DAC),
that asked how this would be done.
As noted in the cover letter, roadmaps do not preclude changes to Tier 2 i
material, they merely ensure that the important analyses and features captured in Tier 1 are adeauately considered in the "50.59-like" change process. They do not create any more procedural constraints than are already called for by Part 52.
NUMARC's letter states that keeping Tier 1 and Tier 2 separate is a
" fundamental" requirement. However, this precept is not supported in practice.
In fact, both the staff and the vendors have agreed that many supporting analyses and details for Tier 1 belong in the Tier 2, and they are currently being put there.
2.
Roadmaps show where severe accidents have been resolved in the design certification rulemaking. They also show where significant policy issues identified in SECY-90-016, " Evolutionary LWR Certification Issues and Their Relationship to Current Regulatory Requirements," and follow-on papers (consolidated into SECY-93-087, " Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced tight-Water Reactor (ALWR)
Designs") have been resolved.
3.
Roadmaps show where the " safety analyses verification" is conducted as part of the ITAAC, as was proposed to the Commission in SECY-91-178 (Concept of ITAAC).
They do this by ensuring that the key features and design basis safety analyses on which the staff based its safety findings are captured in the ITAAC.
4.
Roadmaps can help facilitate the desian certification rulemakinos and minimize extended discussions by documenting how various issues were captured by the ITAAC.
l 5.
Roadmaps will help provide clearer interpretations of the ITAAC and l
minimize confusion when controversies arise durina plant construction.
testina, and in makina the findinos for 52.103 by showing how key Tier 2 information is applicable to and supports the Tier 1 material. This could help both the utility and the NRC staff understand the relative safety significance of issues, and minimize future controversies.
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6.
Roadmaps show where significant issues, USIs/GSIs. lessons learned from operating experience, and key analyses have been resolved in the designs.
7.
Roadmaps can be used by the designers as a front-end discriminator to systematically determine what goes in Tier 1 based on an evaluation of design bases and other analyses.
8.
Roadmaps are very useful to the staff as a " reviewer aid."
They will greatly facilitate the review, and will shorten the time reauired by the staff to make its safety findinas.
Roadmaps aid in verifying that the key design information and insights in the SSAR have beer adequately and accurately translated into the Tier 1 material.
9.
The staff believes that the development of these roadmaps has no sianifi-cant resour,e impacts on the vendors. The roadmaps merely document the development process already followed by the vendors for the ITAAC.
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w FORMAT AND CONTENT OF ROADMAPS In recent meetings, it has become apparent that there is some confusion as to what is meant by "roadmaps." This enclosure explains the purpose of the roacmaps, and states what is expected to constitute a "roadmap."
BACKGROUND Part 52 requires the development of design certification material for the first time. The Tier I design material that is being developed is unique in that the material is presented for the actual systems for the design, while the SSAR and 'ne SRP are more traditionally analysis and engineering discipline oriented. Thus, the development of the Tier 1 information requires the vendors to translate the key elements of each of the SSAR sections to the applicable systems of the facility design. The staff must make its safety findings for the facility design, as well as determine the adequacy of the verification process for the facility (e.g., ITAAC), without an actual plant or the details of the final design.
This is a formidable challenge for both the designers and the staff. There are many andyses addre:, sed in the SSAR that must be addressed in multiple systems of a facility.
For example, analyses for design basis accidents, flooding (internal and external), and fires are addressed in the appropriate section of the SSAR, but must be verified in a final facility by the applicable system ITAAC. This means that each chapter of the SSAR has input to some degree to the development of the Tier I material, and has required a different approach to the development and review of this material than has ever been done before.
In some cases, SSAR analyses and issues are verified by system walkdowns (e.g., equipment qualification).
Others analyses and issues are verified by "boilerplate" ITAAC entries for several individual systems.
Not all analyses in the SSAR are expected to be explicitly verified by an ITAAC, depending on the issues and their safety significance.
In many cases, these analyses in the SSAR are implicitly verified by the ITAAC for the individual systems.
It is not obvious to a third party (staff, public, independent inspectors, and ultimately a utility) how these analyses and issues are addressed in the multiple systems of the ITAAC.
Consequently, the staff has asked the vendors to submit selected "roadmaps" based on their safety significance, showing how certain analyses and issues are addressed in the Tier 1 design material.
The current list of selected "roadmaps" is contained in Enclosure 3.
The staff is examining the ITAAC to ensure that these analyses and issues are addressed adequately.
However, Part 52 places the burden on the vendor to provide ITAAC that are "necessary and sufficient" to provide reasonable assurance that the facility is built and will operate in accordance with the approved design and applicable regulations. The staff's responsibility is to review the ITAAC to ensure that this is done correctly. The roadmaps acument how the analyses and issues are addressed, enable the staff to unGerstand the development and intent of the ITAAC during its review for FDA, and provide a basis for the finding that the ITAAC meet the " reasonable assurance" requirement. The
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w roadmaps also provide to third parties a clear description of how the issues are addressed for use during the design certification rulemaking, as well as t cing future construction, testing and licensing of a facility. These additional considerations are discussed further in Enclosure 1.
FORMAT An example of the format of the cross references from Tier 2 to Tier 1 design
" f: m ti r r "roadmaps" was provided in SECY-92-214, " Development of Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications," of June 11, 1992.
An updated version of the roadmap discussed in ihat paper was submitted by GE in the Stage 3 Design Certifica-tion Material in June 1992, and is attached.
Specific comments on this version of the roadmap are contained in the staff's comments provided to GE in a letter of August 12, 1992. As stated in SECY-92-214, The staff has requested GE to develop a cross reference of key aspects, analyses, and features of the design from the SSAR to the ITAAC in order to document how these issues have been incorporated into the ITAAC.
Specifically, the cross reference will show how key aspects of the accident analyses, PRA, and severe accident issue resolutions are included in ITAAC. contains a preliminary version of this cross reference for containment performance analyses.
For example, the PRA has shown that the vacuum breakers are a key component in analyzing the behavior of the containment for severe accidents.
Thus, the ITAAC for the containment system will include the key facets of the vacuum breakers which are significant contributors to risk for those analyses.
The results of the PRA and severe accident analyses are still being developed, and are expected to reveal many insights into the designs. To ensure the assumptions and insights of these analyses are adequately addressed in the detailed design of a facility, they should be summarized in the SSAR (i.e.,
Chap +.ers 6, 15, and 19).
They also should be reflected in the sections of the SSAR which would be used in the detailed design of the systems for a facility, so that the significance of the insight is clear and readily available for use in the detailed design process. This is a fundamental process that the vendors must go through to complete the documentation of their design in the SSAR, and is discussed in greater detail below.
The vendors should review these assumptions and insights for their significance (based on items such as the relative contribution to risk, sensitivity analyses, operating experience, and engineering judgment). The vendors should then propose a list of leatures that should be verified in the Tier 1 design material, and provide a "roadmap" showing this, as discussed in SECY-92-214.
The staff will review these insights based on their safety significance, and ensure that they are adequately treated in the Tier 1 l
naterial. This process has already been initiated by the staff and the l
vendors.
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Another form of cross reference will exist from the supporting information in the SSAR (i.e., details of the inspections, tests, and analyses for the ITAAC that are being generated) to the appropriate ITAAC that show the appropriate method of satisfying the various ITAAC.
This supporting information will eventually be put in the SSAR, with a reference indicating the appropriate ITAAC it supports.
The supporting information is expected to be put in either Chapter 14 (Verification Programs) or the most applicable sections of the i
SSAR.
Other cross references exist for the information within each of the tiers.
These cross references are not considered "roadmaps" for the purposes of this letter because they do not reference between tiers.
PROBABILISTIC RISK ASSESSMENT (PRA) AND SEVERE ACCIDENT INSPECTIONS. TESTS.
ANALYSES. AND ACCEPTANCE CRITERIA (ITAAC) ASSUMPTIONS AND INSIGHTS The search for assumptions and insights needs to be systematic. This
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includes each vendor re-reading the entire record of the resolution process for the PRA and severe accidents to glean out important insights.
The record includes the PRA, SECY-90-016, " Evolutionary Light Water Reactor Certification Issues and Their Relationships to Current Regulatory Requirements" (and the two Commission papers:
" Issues Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current Regulatory Requirements," and " Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light Water Reactor Designs"), SECY-93-087, " Policy, Technical and Licensing Issues Pertaining to Evolutionary and Advanced Light Water Reactor Designs,"
meeting minutes, faxes on the docket, draft final safety evaluation report 19.2, standard safety analysis report Chapter 19, and formal questions. An example of an important insight is from the advanced boiling water reactor PRA that states that the building housing the diesel-driven fire pump will be constructed in a way that if it collapses during an earthquake, the 4
I failure of the building will not affect the pump's ability to function.
The PRA assumptions and insigtts identified will be " handed off" to the i
combined license (COL) applicant for its use. There is a good chance that few of the people working for a utility (COL applicant) will read all of the PRA to gather assumptions and insights.
By summari. ing and placing the assumptions and insights in one location, the chance fc use and understanding of these insights is increased.
The PRA in and of itself will not be Tier 1 or Tier 2.
Safety assumptions and insights from the PRA (or portions of the insights) will go into the design control document.
Some PRA assumptions and insights may be important enough for inclusion in the certified design (Tier 1). The decision on what goes into Tier 1 l
primarily will be made by engineering judgment.
Assumptions and insights to be considered include the following:
important assumptions in the PRA, structures, systems, and components denoted by importance measures, bypass sequences (containment and suppression pool),
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SFCY-90-016 and SECY-93-087 features, how the design meets containment performance goals, external event, shutdown events, important core damage sequences, what keeps core damage frequency low, and what has large uncertainty and in the extreme could become an important contributor.
Each assumption and insight will include a description of why it is safety significant.
Initial efforts should also identify potential insights that were discarded and why they were.
Examples of why an insight was discarded could be that multiple other systems perform the same function or the numenivoi s eau 1Ls (and implied insight) was caused by overly conservative assumptions or analyses.
In discussions with the vendors, it was decided that for the initial set of insights, if a reviewer had to ponder why an insight was not important, then at a minimum, it was a candidate for an explanation of why it was not included in the insights list.
Next, each vendor proposes which assumptions and insights are to be a
included in Tier 1 and provides a markup of the Tier 1 material. The staff will review the assumptions and insights for completeness and review the rationale for placement in Tier 1 and Tier 2.
The staff also may do independent evaluations to determine which insights should be included in Tier 1.
The staff will provide these insights to the respective appropriate
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technical reviewers or analysts for their consideration, as to whether the insights are candidates for Tier 1.
The staff expects that there will be additional discussions needed between the PRA people and the reviewers / analysts to help the analysts better understand the safety I
significance of pertinent insights.
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4 Table B.1.b: Safety Analysis Verification Using ITAAC SSAR Entry Parameter Value (1)
Verifying ITAAC 6.2.1 Containment Functional Design 6.2.1.1.4.1 Vacuum Breakers Diameter (inches) 20 2.14.1 Primary Containment System Quantity 8
2.14.1 Primary Containment System Table 6.2-2 Drywell 3
Volume (ft )
259,563 2.14.1 Primary Containment System Leak Rate, Drywell and Wetwell WdDay) 0.5 2.14.1 Primary Containment System Wetwell 3
Volume (ft )
210,475 2.14.1 Primary Containment System 3
p Minimum Suppression Pool Water Volume (ft )
126.427 2.14.1 Primary Containment System 2
Total Vent Area (ft )
125 2.14.1 Primary Containment System Vent Centerline Submergence (Low Water Level), (ft):
Top Row 11.48 2.14.1 Primary Containment System Middle Row 15.98 2.14.1 Primary Containment System Bottom Row 20.48 2.14.1 Primary Containment System Table 6.2.2-a RHR System Pump Capacity (gpm/ pump) 4200 2.4.1 Residual Heat Removal System 2
Heat Transfer Area (ft / unit) 2.4.1 Residual Heat Removal System i
Heat Transfer Coefficient (Btu /sec-F) 195 2.4.1 Residual Heat Removal System Service Water Flow (lbm/hr) 2.63x10s-2.4.1 Residual Heat Removal System
$ Table 6.2-2d Secondary Containment
(
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Free Volume (ft )
3.0x10 2.15.10 Reactor Building Pressure (inch H O)
-0.25 2.15.10 Reactor Building 2
Leak Rate (%/ day) 50 2.15.10 Reactor Building l
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w BOADMAP ISSUES I.
SEVERE ACCIDENT DESIGN FEATURES (SECY-90-016/ FOLLOW-UP-PAPERS /
SECY-93-087) (CH. 19.2) 2.
DESIGN BASIS ACCIDENT ANALYSES CORE COOLING (CH. 6)
CONTAINMENT PRESSURE / TEMPERATURE RESPONSE (CH. 6)
RADIOLOGICAL (CH. 15, 19)
OVERPRESSURE PROTECTION (CH. 5)
FLOODING PROTECTION (INTERNAL / EXTERNAL) (CH. 3)
FIRE PROTECTION (CH. 9)
PRA INPUTS / ASSUMPTIONS (CH. 19) 4.
SYSTEMS INTERACTIONS (USI A-17/A-47) 5.
SHUTDOWN RISK 6.
STATION BLACK 0UT 7.
OPERATING EXPERIENCE 8.
TMI/USI/GSI--IMPORTANT FEATURES FOR CLOSURE d
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