ML20056G159
| ML20056G159 | |
| Person / Time | |
|---|---|
| Issue date: | 06/17/1993 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Mcdonald R ELECTRIC POWER RESEARCH INSTITUTE |
| References | |
| PROJECT-669A NUDOCS 9309020158 | |
| Download: ML20056G159 (16) | |
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June 17,1993 Mr. R. Patrick Mcdonald, Vice-Chairman ALWR Utility Steering Committee Electric Power Research Institute 3412 Hillview Avenue Palo Alto, California 94304
Dear Mr. Mcdonald:
I am responding to your letter to Chairman Selin concerning comments of the Advanced Light Water Reactor Program Utility Steering Committee on the staff positions contained in SECY-93-087, " Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs."
In your letter you described areas where you disagreed with parts of the staff recommendations and issues for which you believed written clarification would be useful.
Your letter identified three areas where you disagreed with parts of the staff recommendations. These areas involve the ability of the post-accident sampling system in pressurized water reactors to analyze reactor coolant for dissolved gas, the use of seismic margins methods in the analysis of external events, and the staff's interim position on the reliability assurance program.
Your letter did not contain significant, new information that would cause the staff to change its recommendations in SECY-93-087.
Your letter also requested written clarification for another six issues.
The staff believes that this would not be useful at this time.
Commission approval of the fundamental issues presented in SECY-93-087 is critical to staff progress on the design certification reviews of evolutionary and passive designs. Additional fine-tuning of the staff recommendations at this point will delay these reviews and can more efficiently be handled in the context of the design-specific reviews with only well-defined issues with policy implica-tions being raised to Commission level.
Further, it should be noted that any Commission approval of the staff recommendations in SECY-93-087 would only be tentative and subject to further review in design certification rulemakings.
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Mr. R. Patrick Mcdonald l We agree with your observation on the overwhelming progress that has been mutually achieved in identifying and resolving the technical and policy issues.
related to evolutionary and advanced reactors. We'look forward to working c
with you to finalize the remaining passive policy issues discussed in SECY-93-087 and to successfully complete the design certification reviews.
Sincerely, original signed by Dennis M. Crutchfield, Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation cc:
See next page Distribution:
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Mr. R. Patrick Mcdonald, Vice-Chairman Project No. 669 ALWR Utility Steering Committee EPRI cc:
Mr. John Trotter Nuclear Power Division Electric Power Research Institute Post Office Box 10412 Palo Alto, California 94303 Mr. Brian A. McIntyre, Manager Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Post Office Box 355 Pittsburgh, Pennsylvania 15230 Mr. Joseph Quirk GE Nuclear Energy Mail Code 782 General Electric Company 175 Curtner Avenue San Jose, California 95125 Mr. Stan Ritterbusch Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095
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R. Patrick Mcdonald Advanced Light Water Reactor (ALWR)
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COMMENTS ON SECi 93-087, " POLICY, TECHNICAL, AND Taylor LICENSING ISSUES PERTAINING TO EVOLUTIONARY AND Sniezek ADVANCED LIGHT-WATER REACTOR (ALWR) DESIGNS" Thompson Blaha NRR
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PATRIC MCDONALD AFFILIATION:
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ADDRESSEE:
CHAIRMAN SELIN LETTER DATE:
Jun 7 93 FILE CODE: IDR-14 PT 52
SUBJECT:
COMMENTS ON SECY 93-087, POLICY, TECHNICAL AND LICENSING ISSUES PERTAINING TO EVOLUTIONARY AND ADVANCED LIGHT-WATER REACTOR (ALWR) DESIGNS ACTION:
Appropriate DISTRIBUTION:
CHAIRMAN, COMRS, OGC, SECY, RF SPECIAL HANDLING: NONE CONSTITUENT:
NOTES:
DATE DUE:
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ALWR 7565 [
C /%9 ADW & es CED t ICVf m A *T 9 DfACf09 June 7,1993 Dr. Ivan Selin, Chairman U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Comments on SECY 93-087. " Policy, Technical, and Licensing l
Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs ~
Dear Chairman Selin:
Enclosed are comments of the Utility Steering Committee (USC) of the Advanced Light Water Reactor (ALWR) Program on the subject SECY. We are providing these comments in the spirit of continued cooperation and communication with the NRC staff and Commission. We have been discussing these matters with NRC staff and management for some time and submitted very similar comments in a May 11 letter from the ALWR staff.
We are in agreement with nearly all of the staff recommendations made in i
SECY 93-087. For only three issues does the ALWR Program have points which we feel are " disagreements" with parts of the staff recommendations.
High pressure gas sample in Post Accident Sampling System (#11.1):
We believe this is fundamentally a question of added complexity and cost for very little, if any, net benefit.
Seismic considerations beyond design (#11.N): The utility concern here is the uncertainty in implementation of requirements which go beyond the regime and methodologies where data have been demonstrated to exist. We believe much of the uncertainty here will fall on the future owner during Licensing or operation.
EPRI ALWR Utility Steering Committee 3412 Hillview Avenue, Palo Alto, Califomia 94304
- Telefax: (415) 855-7945 f
w Dr. Ivan Selin June 7,1993 Page 2 Design Reliability Assurance Program (#11.M): We have a process disagreement not a technical one here. We believe the specifics of this program are currently insufficiently understood to warrant inclusion in a rulemaking (" Tier 1").
For another half-dozen issues, we believe some written clarification would be useful.
Intersystem LOCA (#I.F): The staff position in SECY 93-087 has introduced a phrase ("..could not practically be designed...") which seems overly broad in our judgment. We believe the Utility Requirements Document has defined the specific systems for which higher pressure rating is " practical" and NRC staff agreement on those specifics would remove any remaining doubt as to the implications of the staff recommendation.
Common Mode Failure in Digital I & C (#II.Q): Item #4 of staff position is prescriptive and is being read two different ways by industry. This combination of prescription and ambiguity can become a source of regulatory instability.
Multiple Steam Generator Tube Rupture (#II.R): SECY 93-087 talks in terms of making multiple tube rupture a design-basis accident. Recent staff presentations, including at the recent Commission briefing, include clear statements that only " realistic" calculations are intended.
We agree with and require realistic analyses but introduction of new design-basis accident is unwarranted.
Hydrogen Control (#I.G): We accept staff criteria but would like written staff evaluation and confirmation of acceptability of Passive Autocatalytic Recombiners (PARS) so that designers can shift to the control method favored by the utilities.
1.eak Rate Testing (#II.H): This is simply an issue of completeness.
The SECY Policy Issue does not have the same scope as the URD Optimization Issue nor the same (design-related) scope as the previously proposed change to Appendix J. We are not aware of any technical disagreement.
In-service Testing (#I.N): Utility concern here centers on " practicality" and who decides what is practical. This is a difficult judgment and
" conservative" decisions could lead to expensive added equipment (recirculation lines, instrumentation, etc.) and added operational complexity due to extensive at-power testing.
We are continuing to work with the staff on these concerns. We looking forward to briefing the Commission on our views of SECY-93-0S7 both for the sake of specific points mentioned and as a way of continuing the direct dialogue with the Commissioners.
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w Dr. Ivan Selin June 7,1993 Page 3 Finally,let me express my personal thanks for the work the NRC has done over the last several years in reaching advanced reactor policy recommendations and conclusions. The process of early identification and reso!ution of policy and technical issues related to ALWRs is at the center of the entire industry ALWR program and is key to maintaining the nuclear option for this country. We have made real progress.
Sine relvA r
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@LR. Patrick Mcdonald Vice-Chairman, ALWR Utility Steering Committee i
cc Commissioner Kenneth C. Rogers Commissioner James R. Curtiss Commissioner Forrest J. Remick Commissioner E. Gail de Planque Edwin E. Kinter, Chairman USC Dr. Thomas Murley, NRR Joseph Colvin, NUMARC John J. Taylor, EPRI
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l ALWR Program Comments on SECY-93-087," Policy, Technical, and Licensing l
Issues Pertaining to Evolutionary and Advanced Light-Water Reactor ALWR Des igns."
- 1. SECY-90-016 Issues A. Use of a Physically Based Source Term The ALWR Program and the staff are actively engaged in discussions on this issue.
Very recently, changes were made to the Utility Requirements Docu.nent to aid the convergence process between industry and the NRC. We are hopeful of agreement on most,if not all, of the recommendations when they are issued in the near future.
We note that our October 23,1992 letter to the NRC had general agreement with NUREG 1465 except for low volatiles and the need for a regulatory mechanism to allow use of plant specific numbers for certain parameters where the operating plant-based numbers in NUREG 1465 do not apply to ALWRs. The need to allow plant specific numbers applies not only to natural removal mechanisms (as noted in SECY-93-OS7), but also to other source term parameters such as fission product release timing and gap release magnitude.
It is also noted that the differences in low volatile release fractions are due mainly to experimental data on low volatile release (both in-vessel and ex-vessel). Contrary to the staff position in the SECY, we believe that cavity flooding, while expected to reduce ex-vessellow volatile release,is not the primary reason for the difference between the ALWR numbers and NUREG 1465.
D. Station Blackout As noted in SECY-93-087, we continue to believe this issue is not applicable to the passive plant designs since there is a fundamentally different, and greatly diminished safety role for ac power in the passive plant concept.
F. Intersystem Loss-of-Coolant Accident i
The ALWR Utility Requirements Document has always sought resolution of this issue via increased pressure design for connected systems. The clarification made to the Requirements in May of 1992 were intended, among other things, to provide uniform guidance on practicality for this increased design pressure resolution. This has been thought necessary because there is some indication that the staff will require the added monitoring and testing of isolation valves independent of the pressure rating of the connected system. We believe only a clarification is necessary
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ALWR Program Comments on SECY-93-OS7 from the NRC for the phrase "could not practically be designed to meet such a (URS pressure) criterion."
G. Hydrogen Control The URD requires that designs meet the hydrogen generation and hydrogen concentration criteria specified by the staff. However recent URD changes have been submitted to endorse a different type of hydrogen control system. The new type of system is called a Passive Autocatalytic Recombiner (PAR). Supporting technical material on one design of a PAR developed in Germany has been submitted to the staff. We believe this technology should be allowed in place of electrically-powered igniters and thermal recombiners.
N. Inservice Testing of Pumps and Valves We have no disagreement with the staff on adequate testing of safety-related pumps and valves. We continue to believe the ASME Section XI can provide the appropriate guidance in this area. We encourage the NRC to work with the ASME code committee to develop an industry code whichis mutually acceptable. On the issue of " practical design" for testing, we continue to believe considerations of practicality must be encouraged and clarified. Such encouragement should specify avoidance of excessive complexity in design due to test features. Practicality considerations should factor in cost-effectiveness for both design and operation.
II. Other Evolutionary and Passive Design Issues B. Electrical Distribution We continue to believe this issue is not applicable to the passive designs but do not disagree that the staff resolution can be made in the context of the regulatory treatment of non-safety systems H. Containment Leak Rate Testing We agree with the staff recommendation onType C testing. However, our i
Requirements have also noted other differences with containment testing between i
ANSI /ANS 56.8 and Appendix J which we also believe should be resolved for ALWRs. This may merely require a clarification whether or not these are in fact
" policy" concerns since we believe there are no technical differences between us and the staff.
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ALWR Program Comments on SECY-93-OS7 L Post Accident Sampling System Despite generally fruitful discussions with NRC staff on this issue, we continue to disagree on the need for a pressurized (gas) sample for post-accident analysis.
Neither deterministic nor probabilistic safety analyses have identified any significant accident scenarios where such samples would be useful. While the staff has recognized this for the boiling water reactors, we continue to believe it is equally true for advanced pressurized water reactors. It is important to note there is a significant design difference between requiring a high pressure (gas) sample and requiring a (low pressure -liquid) sample from a high pressure system. Sample volume, system valving, potential for spills outside containment are all aspects of that design difference which we believe argue in favor of the system described in the ALWR Requirements.
In any event, the SECY-93-087 description of the need for chloride sampling is unclear both calling it "not... mandatory" and "reauired" We believe this point should be clarified.
M. Reliability Assurance Program The ALWR Program agrees with the staff that the Reliability Assurance Program (RAF) can be divided into the two stages, design (D-RAP) and operations (0-RAP),
j as described in SECY 93-087.
Regarding the D-RAP, the ALWR Program does not believe that it is necessary to include in design certification a Tier 1 description of D-RAP. This process is described in adequate detailin the applicant's SAR to allow the staff to review and conclude with a high level of confidence that a D-RAF with appropriate attributes will be implemented during the design process. We would emphasize the specific objections to inclusion in Tier 1 that NUMARC has raised in their response to SECY-93-087. Additionally, the ALWR Program recommends that the staff close this portion of the RAP issue for both the evolutionary and passive designs by endorsing the current requirements set forth in the URD.
Regarding the O-RAP, the ALWR Program believes that for both the evolutionary and passive designs," Life of the Plant" reliability considerations will be adequately addressed via standardization activities consistent with the industry approach to implementation of the Maintenance Rule. For passive designs these activities will include performance and condition monitoring and related requirements for SSC's identified as a result of the regulatory treatment of nonsafety systems (RTNSS) process. It should also be noted that as the majority of these activities will be based on the requirements of the Maintenance Rule (10 CFR 50.65) that they are expected to be integrated into the COL applicant's Maintenance Program.
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ALWR Program Comments on SECY-93-OS7 N. Site-Specific Probabilistic Risk Assessments and Analysi. of External Events We have significant concerns regarding key aspects of the NRC staff position on the performance of seismic margin assessments (ShiA) for demonstrating ALWR seismic design capacity beyond the design basis. Our principal concern relates to the proposed seismic margin earthquake (ShiE) level of two times the ALWR design basis earthquake which results in an ShiE of 0.6g peak ground acceleration (pga). In addition, we are concerned about the proposed use of "PRA-based" methodology to the exclusion of other accepted methods.
PrincipalIndustrv Concern: Seismic hiartin Fgrthquake level The NRC staff proposal for an ShfE of two times the design basis earthquake is arbitrarily excessive and unnecessary for demonstrating seismic capacity beyond the design basis. A seismic margin of 1.67 times the design basis earthquake is specified in the Utility Requirements Document (URD), not 1.5 as reported in SECY-93-087; this results in an SNIE at 0.5g compared to the design basis earthquake of 0.3g pga.
Contrary to NRC staff remarks at the hiay 14 Commission briefing on SECY-93-087, the seismic margin demonstration required for ALWRs by the URD is significantly more stringent than that ren sted of operating reacto:s as part of the Individual Plant Examination of Externa; Events (IPEEE) for severe accident vulnerabilities (Generic letter 88-20, Supplement 4). This is evidenced by the fact that operating plants with design basis earthquakes up to 0.2g are permitted to utilize an Sh1E of 0.3g (i.e., demonstrate a margin of 1.5 times the design basis earthquake). yet. for an ALWR located at the same sites. the URD required ShiE would be 0 Sg.
It shouM be noted that most candidate sites for advanced plants in the Central and Eastern United States are expected to have site-specific SSEs substantially below the standardized design basis earthquake of 0.3g pga (indeed, most sites are likely to have an SSE at or below 0.2g). Thus, a 0.5g ShiE is expected to demonstrate a seismic margin factor for an as-constructed ALWR significantly greater than 1.67.
From a seismic risk persoective. it is the marcin above the site-specific ground motion (as determined in accordance with 10 CFR Part 100. Appendix Bt not the marcin relative to the desien basis earthouake. that is relevant. The URD required ShiE of 0.5g provides ample margin in seismic capacity relative to the site-specific SSE for most candidate ALWR sites and is significantly more stringent than the Sh1E requirement for like-sited operating plants. For these reasons, we believe there is not adequate technical basis for requiring an ShiE greater than 0.5g.
In developing the URD requirements for seismic margins analysis, the ALWR Program has been mindful of the possible future implications. We believe that an ShiE requirement of two times the design SSE would introduce significant practical complications and entail substantially more engineering effort for the design certification applicant and the COL holder - without a correspondinc safety benefit.
For example, while not required to address any actual seismic vulnerability, the page 4
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industry has determined that the proposed 0.6g ShiE criterion - not the 0.3g design basis earthquake - may inappropriately govern the design of the plant in some areas.
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We believe that an Sh1E at two times the design basis earthquake represents a potential for significant increase in the level of engineering effort and component cost for analysis at the desiga certification stage and procurement at the COL stage in crder to demonstrate seismic margin. hiost important is the fact that while earthquake experience data up to a level of 0.5g has been accumulated, significantly less data exists for components at higher acceleration levels. Thus, additional data will have to be generated to analytically demonstrate the incrementally greater seismic margin sought by the NRC staff. A requirement to demonstrate seismic capacity above 0.5g will therefore adversely impact the procurement of equipment by limiting the availability and increasing the cost of hardware where there is i
uncertainty about demonstrating comp 6nent ruggedness at this overly conservative level.
-r We take it as evidence of the robustness of ALWR seismic designs that the staff has observed that at least one of the designs under review may be able to demonstrate a seismic capacity of two times the 0.3g design basis earthquake. However, the industry is concerned that the NRC staff's design review of a single ALWR design at this stage of design development provides inadequate technical basis for their i
apparent conclusion that 0.6g is an appropriate and practical ShiE requirement. As a
noted above, the NRC staff has not adequately considered the future implications l
and workability of a 0.6g ShiE relative to detailed design or verification of procured I
and constructed structures, systems, and components.
AdditionalIndustry Concern: Acceptable ShfA hiethodologies i
A related industry concern is the methodology to be used for demonstrating ALWR seismic capacity beyond the design basis. The NRC staff position indicates that a "PRA-based" seismic margins approach is preferable to a seismic PRA given the l
unsettled state of seismic hazards methodology. We agree that a seismic margins approach is preferable, and we believe that the use of the EPRI seismic margins methodology approved for use with operating plant IPEEEs and specified in the URD (i.e., assuring beyond design basis seismic capability of two complete safe shutdown paths) should be permitted for seismic assessment of advanced plants in lieu of requiring the ~PRA-based" methodology proposed by the staff.
l Recuested Commission Actions Recardine SMA i
In summary, the Commission should direct the staff, in response to SECY-93-087, to adopt the URD requirement of 0.5g for the SME. That value assures adequate j
seismic capacity beyond the design basis, especially considering that, from a seismic risk perspective,it is the mr.rgin above the site-specific ground motion (as determined in accordance with 10 CFR Part 100, Appendix B), not the margin P3Se 3 J
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relative to the design SSE, that is relevant. Likewise, the 0.5g SME requirement l
specifkd in the URD will not unduly burden the industry as would the higher SME proposed by the NRC staff. We trust that the Commission will conclude as the i
industry has that the significant additional burden and complexity introduced by a higher SME requirement would not be offset by a corresponding increase in plant safety.
The Commission should also direct the staff that either "PRA-based" or the existing IPEEE and URD seismic margins methodologies are suitable for advance LWRs.
Q. Defense Against Common-Mode Failures in DigitalInstrumentation and f
Control Systems Item # 4 of the revised staff position indicates the specific need for a set of safety-l grade displays and controls. The rationale for requiring these controls and displays i
be safety-grade and the design basis for which they are to be safety qualified is not understooi If the assumption is that plant safety and nonsafety capabilities to actuate needed functions per a 1-3 is lost, then the basis for controls / displays in # 4 would be that they are available given some very widespread and catastrophic event that disables many separate and diverse features. Depending on the specific architecture, design requirements and regulatory oversight, there are backup l
hardwired nonsafety capabilities for defense against common-mode failures that are i
as adequate as that indicated in item # 4 and which should be allowed.
1 In the second paragraph of item 4 there are sentences that appear to be in conflict 1
with each other. The first sentence of this paragraph says," to the lowest level practical". Latter portions of the paragraph are very specific about how and where the hardwires are to be connected to the architecture and these are not always practical. It is suggested that the specificity of this wording be removed and i
replaced with wording at a functionallevel, to require that these controls and displays not be subject to the common mode failure identified in items 1 and 3 above.
R. Steam Generator Tube Ruptures j
We continue to believe that the desien basis steam generator tube rupture (SGTR) event for ALWRs should be the single tube rupture; however, we concur with the staff that designers of ALWRs with depressurization systems consider multiple l
SGTR event scenarios as part of the plant Probabilistic Risk Assessment. For that t
reason, the URD has been revised to require that the Plant Designer analyze the plant response, on a best estimate (BE) basis, for a spectrum of multiple SGTRs.
This requirement assures that the Plant Designer explicitly determine the plant j
response to a range of multiple SGTRs and determine the contribution to core damage and release frequencies. These analyses will quantify the plant transient behavior considering the potential for depressurization of the primary system. The page 6 1
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analyses shallinclude assessment of the thermal-hydraulic and neutronic behavior of the system during SGTR events.
Based on preliminary reviews, Westinghouse and the ALWR Program expect more detailed analyses of multiple SGTRs (involving up to five tubes) to show that the AP600 can respond to these events without actuating the depressu ization system.
The test program for the AP600 includes tests to develop the nece. sary data for models used for MSGTR analyses. It is ex oected that the experimental data from the planned testing program will provide an adequate basis for evaluating the response of the AP600 to a multiple SGTR event.
T. Control Room Annunciator (Alarm) Reliability We have no comment on this issue except that quotes from Chapter 10 of the URD do not reflect the latest revisions.
III. Issues Limited to Passive Designs A. Regulatory Treatment of Non-Safety Systems in Passive Designs The discussions between industry and the staff on this very difficult issue are still very active. A g. eat deal of progress has been made and we look forward to reaching a mutually satisfactory resolution.
E. Control Room Habitability In SECY-93-OS7, the staff expressed serious reservations concerning the feasibility of the control room pressurization system. We believe implementation of the Requirements for such a system is practical including those Requirements calling for systems that can be periodically tested. Dialog continues with the staff in connection with the new accident source term and we believe those discussions will go a long way toward ameliorating NRC staff reservations.
F. Radionuclide Attenuation We are actively engaged in discussions with the staff on this issue and expect to reach complete or close agreement on this and other aspects of the Source Term issue.
G. Simplification of Offsite Emergency Planning We have recently submitted URD changes in this area. We expect additional supporting material will only become available after the issuance of the Safety page 7
w ALWR Program Comments on SECY 93-087 Evaluation Report for Volume III of the URD and therefore we anticipate the need for a SER Supplement once all material has been submitted.
As stated in the recently submitted URD revisions, the technical basis for changing emergency planning is the technical design criteria together with the design certification demonstration that the criteria are met. Thus, while we agree with the staff statement in SECY-93-087 that detailed design information is necessary to fully resch e the issue,it is appropriate that staff evaluation of the criteria proceed at this time in order to define what standards the detailed design must meet.
Also, we agree that very low calculated probability values is not necessarily sufficient basis for emergency planning change. This is why we have proposed deterministic criteria with PRA in a complementary role, as described in the NRC Severe Accident Policy Statement.
Finally, while we have developed the technical basis for emergency planning change in the context of the passive plants, we believe as stated in the forwarding letter for our URD revisions that any ALWR, including evolutionary plants, should have the opportunity to demonstrate that it meets the criteria and thus be considered for ALWR emergency planning.
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