ML20197F173
| ML20197F173 | |
| Person / Time | |
|---|---|
| Issue date: | 03/20/1986 |
| From: | Bilhorn S NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Newton C ENERGY, DEPT. OF |
| References | |
| REF-WM-1 NUDOCS 8605150463 | |
| Download: ML20197F173 (1) | |
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% 20 JLinehan NO RRBoyle Carl Newton Office of Geologic Repositories SCoplan Office of Civilian Radioactive JKennedy Waste Management RCook U. S. Department of Energy PPrestholt GB-270 Tverma Washington, DC 20585 JGiarratana SBilhorn & r/f
Dear Mr. Newton:
PHildenbrand RJohnson Enclosed are copies of U. S. Nuclear Regulatory Commission (NRC) response to Department of Energy Q-list questions and NRC response to the State of Nevada questions stemming from the December 1985 Quality Assurance (QA) meeting.
Since both of these documents provide follow-on information from.the December QA meeting, I have distributed them to the attendees of that meeting as marked on the attached list.
Please distribute to those remaining, as appropriate.
Thank you.
Sincerely, Susan G. Bilhorn Repository Projects Branch Division of Waste Management Office of Nuclear Material Safety and Safeguards
Enclosures:
As stated.
WM Rec d File WM Project 05 Docket No.
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Mr. James Knight Director Licensing and Regulatory Division Office of Geologic Repositories U. S. Department of Energy RW-20 hashington, DC 20585
Dear Mr. Knight:
Prior to the U. S. Department of Energy /U. S. Nuclear Regulatory Commission (DOE /NRC) Quality Assurance (QA) meeting December 4-5, 1985, the 00E provided the NRC with a series of 13 questions referencing " implementation of Q-list methodology".
In the minutes of this meeting NRC Staff committed to sending formal responses to each question. The purpose of this letter is to transmit preliminary responses to these questions to the 00E. The subjects addressed are complex and will require additional interaction between our staffs.
The information contained in these responses is therefore preliminary and intended to provide a basis for discussion between our staffs.
During development of these responses a number of the questions were subject to interpretation.
In these cases the response has been directed to address what appeared to be the underlying concern.
In addition, some questions suggest a misunderstanding of the legal constraints associated with NRC regulations.
Should you have any questions concerning these responses, please feel free to contact S. Bilhorn of my staff (FTS 427-4682).
In addition to these responses, the staff is in the process of developing a draft generic technical position paper (GTP) on the methcdology for determining what items and activities are important to safety and important to waste isolation. The draft positions were summarized in the December QA meeting and i
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405/58/86/02/07 2-are contained in the minutes to that meeting. The GTP is scheduled for publication as a draft document within the next 6 months.
Sincerely, John J. Linehan, Acting Chief Repository Projects Branch Division of Waste Management Office of Nuclear Material Safety and Safeguards
Enclosure:
Response to DOE Questions on Implementation of Q-list Methodology
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Response to DOE Questions on Implementation of Q-List Methodology 1.
Allowable Dose Criteria (Operations Phase) 4 1.1 Question - 60.111(a) and 60.2 indicate that 0.5 rem is the threshold
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value for making a determination on "important to safety." Should this be considered an upper level accident dose limit as well?
The staff response to this question has been combined with response to question 1.2 below.
1.2 Question - What should the dose limit for design basis accidents be?
10 CFR 60.111(a) states that "the geologic repository operations area shall be designed so that until permanent closure has been completed, radiation exposures and radiation levels, and releases of radioactive materials to the unrestricted area, will at all times be maintained within the limits specified in Part 20 of this chapter and such generally applicable environmental standards for radioactivity as may have been established by the Environmental 4
i Protection Agency." The applicable EPA standard, 40 CFR Part 191, requires reasonable assurance that the combined annual dose equivalent to any member of the public in the accessible environment not exceed 25 mrem to the whole body, 75 mrem to thyroid, and 25 mrem to any other critical organ (40CFR191.15).
10 CFR Part 20 places the annual whole body dose limit to any individual in j
unrestricted areas at 0.5 rem (10CFR20.105).
10 CFR 60.2 establishes 0.5 rem as he threshold value for determining what systems, structures and components are "important to safety" in order to ensure that those items whose failure could lead to higher exposures will function as required. Therefore the design bases for the period before permanent closure should consider the off site dose limit for an accident as 0.5 rem.
l In the context of licensing other types of facilities, the NRC has defined
" design basis accidents" as those accidents whose likelihood of occurrence is deemed to be credible and for which engineering safety features assure that public health and safety will not be endangered.
For these other facilities, protection of public health and safety involves the identification of the credible accidents against which the design of the facility will be tested.
After identifying the credible accident scenarios, the potential consequences of the design basis accidents are then evaluated to determine whether the predicted consequences fall within the appropriate dose guidelines. The l
purpose of the design basis accident and the associated dose guidelines has been to test the facility design to determine if the safety features can adequately cope with accidents, and to evaluate the suitability of the proposed i,
site.
In addition, past reactor licensing practice has used the accident dose guidelines as one of the criteria for determining what equipment was " safety related," and therefore subject to 10 CFR Part 50, Appendix B.
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1 Unlike the reactor site criteria in 10 CFR Part 100.11, or the independent spent fuel storage installation (ISFSI) criteria in 10 CFR Part 72.68, 10 CFR Part 60 does not specifically refer to a " design basis accident" and does not J
explicitly establish pre-closure accident dose guidelines. However, 10 CFR j
Part 60 does specify a dose limit for determining what items will be "important i
to safety". The term "important to safety" in 10 CFR Part 60 is used to j
determine what pre-closure items should be on the Q-list.
The rationale behind j
placing a system, structure or component on the 0-list is to assure, via application of additional QA and design requirements (10CFR60.152 and 10CFR60,131(b) respectively), that it will perform its intended function.
Establishing a design basis accident dose limit higher than 0.5 rem would not be consistent with the dose limit specified in the 10 CFR 60.2 definition of important to safety.
1.3 Question-60.2 states in part that "... engineered structures, systems and components essential to the prevention or mitigation of an accident that could result in a radiation dose of 0.5 rem or greater..." are important to safety.
In light of questions 1.1 and 1.2, should mitigative systems be deleted from that definition?
As noted in response to Questions 1.1 and 1.2, 10CFR 60.111(a) requires systems, structures and components to be designed to maintain the dose to the unrestricted area to 10 CFR Part 20 limits. The 00E should note that the object of the "important to safety" definition in 10 CFR 60.2 is to provide assurance that the 0.5 rem dose is not exceeded during pre-closure accidents.
Those systems, structures and components essential to mitigate doses to the 0.5 rem level are considered important to safety to assure, through QA, design and other applicable requirements, that they will perfo m their safety function.
1.4 Question - The 0.5 rem threshold dose is based on the permissible annual dose to the off-site population resulting from normal operation, as defined in 10 CFR 20.
If 10 CFR 20 is revised, will the 0.5 rem threshold also be revised? Can we interpret the 0.5 rem dose as a whole body equivalent dose?
The staff thinks it is important to stress that the 0.5 rem dose limit referenced in 10 CFR Part 20 is not the permissable annual oose to the off site population resulting from normal operation, but rather the maximum annual dose to an individual in the unrestricted area. DOE should consider the EPA standard of 25 mrem to the whole body, 75 mrem to thyroid, and 25 mrem to any other critical organ as the permissable annual doses to the off site population resulting from normal operation (40CFR191.15).
The Supplementary Information accompanying the proposed revisions to 10 CFR Part 20 indicates that the NRC will update other parts of its regulations after the revisions to 10 CFR Part 20 become final.
The staff anticipates that the 0.5 rem figure in the definition of "important to safety" will be considered in that update.
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As stated in the Supplementary Information to this proposed revision, the NRC staff considers the " effective whole body dose equivalent" concept to be more technically appropriate than the " dose to the whole body, or any organ" concept currently used in defining "important to safety" (10CFR60.2). For this reason, j
the staff agrees that the "whole bcdy dose equivalent" concept may be used pending resolution of the 10 CFR Part 20 rulemaking.
2.
Analytical Assumptions (Operations Phase) 2.1 Question - When design details are lacking, what is an acceptable j
basis for estimating dose consequences of design basis accidents?
A primary objective at this stage of the repository program should be to determine the specific functions and functional requirements of a structure, system, or component and to identify scenarios which may exceed the functional recuirements. The information obtained from this analysis should then be applied to determine what design details are necessary to assure that the j
requirements will be met.
As noted in response to Questions 1.1 and 1.2, 10 CFR Part 60 does not 2
explicitly address design basis accidents. The following response therefore addresses the question restated as follows: At the early stage of this first-of-a-kind program, when design is in a conceptual phase but work is ongoing, what is an acceptable method and information base for estimating dose consequences of accidents?
1 The staff acknowledges that this is a difficult task based on the limited l
information available upon which to base major decisions. Accident scenerlos including initiating events as well as dose consequences for accidents w1'l need to be identified and estimated based on conservative engineering judgment and existing information. The available information base may include data 4
collected and analyzed for other similar activities, such as external events 4
i for reactor facilities and design basis accidents for ISFSI's and refuelbig operations at nuclear power plants where these can be shown to apply direttly to the HLW facility. Although the respository operational system represents a unique nuclear facility, perhaps correlations can be made with other similar nuclear facilities in order to enable knowledgeable decisions to be made and to avoid repetition of effort and prior mistakes.
Extrapolation of analyses conducted with analogous facilities must be carefully conducted and the information obtained rigorously examined to assure that key differences in facilities have not been overlooked. Many factors need to be taken into account when estimating the consequences of an accident and the potential dose to the unrestricted area. These factors include release rate.
source term, meterologic conditions at the site, and location of release.
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2.2 Question - Part 60 contains numerous references to " credible" events to be considered in design. What is an appropriate definition of credible event?
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1 The term " credible" with reference to events to be considered in design is used in 10 CFR 60.130(b)(3) " credible fires or explosions" and 10 CFR 60.133(a)(2)
" credible disruptive events, such as flooding, fires and explosions". The former is a criterion for the geologic repository operations area..while the i
latter is a criterion for the underground facility. As noted in the Supplementary Information to the rulemaking establishing the technical criteria i
in 10 CFR Part 60, the design criterion pertaining to continued operation l
during and after fires has been limited to such events as are " credible." This revision was made in response to comments that suggested that the proposed 1
language corld be interpreted to require protection against any fire or i
explosien that might be physically possible. 48 Federal Register 28194, 28213, j
June 21, 1983.
9 Events, internal or external to the HLW facilities, are initiators of accident scenarios.
Internal events, such as equipment malfunction or cperator error, j
are direct initiators while external events, such as floods or earthquakes, are indirect initiators that may result in an internal event which then initiates an accident scenario.
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The term " credible event" would refer to that event which is sufficiently likely to warrant consideration ir. design of the facility in order to prevent J
or mitigate the consequences of their occurrence.
2.3 Question - What is an appropriately conservative probability value for credible events / accident scenarios?
I' The intent of equating credible events with accident scenarios is unclear.
Accident scenarios include the initiating event, all related common mode failures and any additional independent failures, and release and transport of radionuclides to the unrestricted area.
Events, defined as above, are potential initiators of accident scenarios.
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The following response addresses a similar question previously posed by 00E:
l What does the staff consider an appropriate lower probability limit for j
accident scenarios considered in the design basis.
It is the staff's position that credible initiating accidents should not be i
bound by a specific probabil',ty value at this stage in the repository program.
It is important to note that for new types of facilities where it may be difficult to evaluate the safety of the facility due to limited experience with 1
I the technology, the NRC has factored extremely low probability, high consequence events into their evaluation of the facility.
For example, because j
of the difference in technology and experience between the Clinch River Breeder i
Reactor and a typical light water reactor, additional measures were required for Clinch River against accidents beyond the established design basis.
It should also be emphasized that in terms of reactor licensing requirements and i
analysis, probability has generally not been used to identify design basis accidents. The staff expects to follow the same general approach in reviewing a repository license application.
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5 2.4 Question - At what accident scenario probability value should the 5 rem limit proposed in the response to question 1.2 apply?
Response to this question is not appropriate as the 5 rem level proposed by DOE has not been accepted (see response to Questions 1.1 and 1.2).
See also response to Question 2.3 for discussion regarding establishment of probability based values.
2.5 Question - The Commission's Part 60 rule at various places alludes to the need for... redundant systems to the extent necessary to maintain...
the ability to perform their safety functions" (e.g. 60.131(b)(5)(ii))
(emphasis added).
In other places, the rule specifies redundancy as in 60.131(b)(10)(iv), "...shall be designed to include two independent indicators..." Does the Commission intend there to be a uniform rule on redundancy and therefore the necessity to design for independent single failure?
The Commission does not reouire redundancy except as specified in 10 CFR Part
- 60. The rule is to design to ensure that the continued function of the equipment is retained. Redundant equipment should be employed where necessary and apprcpriate. Single failure of components which result in loss of capability of systems to perform independent safety functions should be analyzed. Where necessary to assure the dose limit is not exceeded, systems must be designed to address independent single failures.
2.6 Question - For structures, systems and components whose failure to perform their intended function could result in an accident resulting in a dose commitment greater than 0.5 rem, can the accident be precluded by design or will non-mechanistic failures be imposed?
The staff supports the concept that non-mechanistic failures should not be imposed as a design condition if, via analysis, the failure of those structures, systems, or components can be demonstrated not to exceed the dose limit to the unrestricted area. The purpose of placing a system, structure or component on the Q-list is to assure, via application of additional design and QA requirements, that it will perform its intended function.
3.
Waste Isolation 3.1 Question - What criterion should be used to define Important to Waste Isolation?
The term " isolation" is defined in 10 CFR Part 60 as:
" inhibiting the transport of radioactive material so that amounts and concentrations of this material entering the accessible environment will be kept within prescribed limits." Based on this definition, and the performance c5jectives of 10 CFR 60 Subpart E, the term " barriers important to waste isolation" (10CFR60.151) means those natural or engineered barriers that contribute to meeting the containment and isolation requirements of 10 CFR Part 60.
10 CFR Part 60 references 40 CFR
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Part 191, the Environmental Protection Agency (EPA) standard for overall repository system performance.
The items and activities important to waste isolation will be dependent upon what barriers are relied on to meet the performance objectives of 10 CFR Part 60 and will include:
A.
Components of the engineered barrier system (waste package and underground facility),
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Components of the natural barrier system, C.
Items and activities necessary to support the determination of l
whether the performance objectives will be met, 4
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Items and activities whose behavior could significantly degrade postclosure performance, and l
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Items and activities important to safety that could affect postclosure performance.
i 3.2 Question - Should systems, structures, and components important to waste isolation be included on the Q-list?
Yes. As stated in staff comments 9 and 11 from the December 4-5, 1985 quality assurance meeting minutes, structures, systems and components important to waste isolation and certain activities should be included on the Q-list.
l 3.3 Question - Will the NRC require the application of the single failure criterion to repository facilities prior to closure?
As this question relates to preclosure and single failure, it has been addressed in response to Question 2.5 above.
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I 405/SB/86/02/09 1 - l'AF, J7 33 i
i Mr. Robert Loux Director i
Nuclear Waste Project Office 4
State of Nevada i
Capitol Complex Carson City, NV 89710
Dear Mr. Loux:
Several coments were made by the State of Nevada following review of the i
minutes for the U. S. Department of Energy /U. S. Nuclear Regulatory Comission (DOE /NRC) meeting on Quality Assurance, December 4-5, 1985 (letter from R. Loux to R. Browning, 1/22/86). Enclosed is the staff's response to those coments i
and concerns.
a We appreciate your coments and the opportunity to address the States' concerns as they arise.jp responding to these comments we acknowledged a basic agreement between the State of Nevada and NRC staff on the issues addressed. The staff realizes that having not been able to attend the subject meeting, the State may-have found 't difficult to interpret the content of a meeting based on sumaries pres 9ted in the meeting minutes. We would like to note, however, that these meeting minutes represent, observations which should be read in the context of the specific areas of quality assurance discussed in the meeting.
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We hope the enclosed responses provide adequate information.
If you have any questions or coments please feel fret to contact J. Linehan, Acting Chief of i
the Repository Projects Branch at (30,1) 427-4177.
W g g' Robert E. Browning, Director p
Division of Waste Management
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Office of Nuclear Material Safety 1
and Safeguards 1
Enclosures:
i 1.
NRC Staff's Response to State of Nevada Comments in Letter to i
R. Browning, 01/22/86.
2.
Minutes from DOE /NRC QA Meeting, December 4-5, 1985.
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3.
Preliminary Draft NRC "Q-list" Positions Presented December 5, 1985 l
4.
Letter from NRC to DOE - Level of i
Detail in SCP, December 12, 1985.
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- j Response to State of Nevada Concerns and Comments from 01/22/86 Letter
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The NRC's role with respect to QA on non-Q-list items and pre-site characterization activities.
i The NRC is committed to the " ideals of quality assurance".and plans to assess all areas of the DOE program related to assuring that a geologic repository will function as required to protect public health and safety. We think our l
commitment is demonstrated in the level of effort expended by NRC staff to
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identify and provide guidance on QA issues through the QA and technical's in eractions conducted with 00E over. the past several years. We acknowledge that 00E observation number 1 in the minutes from the December 1985 QA meeting (Enclosure 3) may somewhat obscure the intent of our overview of non-Q-list items and activities. However, staff' positions on the Q-list presented during 1
the same meeting (Enclosure 3; 1.1,1.2 and 1.3) provide more detail in defining our oversight responsiblities.
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The Q-list is comprised of those items and activities that, due to their importance to safety and to waste isolation, need high levels of assurance to prove that a repository can operate as required. However, the staff will review all items and activities necessary to meet the licensing requirements 1
and support a license application. Any information necessary to demonstrate compliance with these requirements must have ' adequate assurance of quality.
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Those items and activities that are determined by DOE not to be on the Q-list j
or not to be used or referenced in the license application will also be
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assessed by the staff to evaluate the adequacy of that determination.
j NRC's recommendation in the Ford Amendment Study (NUREG-1055), referenced by Nevada, addresses an issue unique to reactor licensing and not applicable to the repository program.
In reference to quality assurance in the reactor program, there are two classes of safety items - safety-related and important i
to safety - that have been referenced in the regulations. The distinction between and requirements associated with these two classes has been the cause j
of much discussion.
Important to safety is a broa.ier class of items that includes items necessary to meet the statutory requirements of providing reasonable assurance that the facility can operate without undue risk to public i
health and safety. Items classified as safety-related are a subset of those important to safety and require the use of 10 CFR 50, Appendix B QA requirements while items important to safety can use lesser QA measures.
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2 The NRC has taken a conservative approach in the repository program by requiring that all items and activities that are important to safety or important to waste isolation be subject to 10 CFR 50, Appendix B QA requirements and NRC review. As indicated above, anything necessary to demonstrate compliance with the requirements in 10 CFR 60 will be reviewed by the staff and must be supported by adequate information.
The staff has considered the results of the Ford Amendment Study (NUREG-1055) and is using the appropriate recommendations in the repository program. We consider this a valuable source of information as it represents a unique review of lessons learned in quality assurance through the experience in the nuclear power reactor program. To enhance our use of this experience, we have involved two of the authors of NUREG-1055 in the repository QA activities. Both were key participants in the December 1985 DOE /NRC QA meeting.
The State of Nevada also expressed a concern in this comment that the NRC will not give " proper QA scrutiny" to pre-site characterization activities. The NRC has continually stated that all information to be used or referenced in the license application will need to have adequate assurance of quality and will be-j reviewed by the staff. This includes, of course, pre-site characterization activities that will be used or referenced to support licensing findings. This is clearly addressed in the staff position 1.4 on the Q-list as presented in the December 1985 OA meeting (Enclosure 3). The staff has to date been involved in review and evaluation of pre-site characterization activities through participation in numerous QA and technical interactions including workshops, data reviews and other DOE /NRC meetings on generic issues.
Qualification of existing data, that is pre-site characterization data collected before implementation of the QA program, is a subject of concern and the focus of much discussion. The staff presented a summary of the draft staff generic technical position (GTP) on qualification of existing data in the QA meeting (see Enclosure 12 to the meeting minutes). This draft GTP is scheduled to be released for public connent within the next several months.
In addition to oversight of the DOE activities, it is important to note that the NRC will maintain awareness of the States' testing programs, insofar as to evaluate whether DOE's information is adequate and that the States' programs do not adversely impact the site.
In the NRC review of the DOE site characterization program, the NRC will evaluate whether DOE has taken into account the information from and potential impacts of the States' testing programs. We think it is necessary to note that a consideration in NRC's evaluation of information collected by the States will be the extent to which that information is supported by an adequate QA program.
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The concern raised in the State's second comment is related to the DOE Observation 2 in the minutes from the December 1985 QA meeting. NRC Observatte, 11 and staff position 2.2 in the (Enclosures 2 and 3) addresses the staff's position on this issue in the context of the QA meeting. The main purpose of the Q-list is to provide a general listing of the items and activit'es which fall under the 10 CFR 50, Appendix B QA program.
It is a starting point for the staff's review of the scope of the QA program. The staff will utilize information in other sections of the SCP, such as in section 8.3 and its' references, to review the scope of the test program in detail.
The staff's guidance with respect to the content of the SCP are addressed in Draft Reg. Guide 4.17 " Standard Format and Content of Site Characterization Plans for High-Level-Waste Geologic Repositories".
Staff guidance and discussion, which basically agree with your comments, were most recently presented in the DOE /NRC meeting on October 29-30, 1985 and in the December 12, 1985 letter from NRC to DOE addressing the level of detail expected in section 8.3 of the SCP (Enclosure 4).
The staff believes that the DOE's QA program is an integral part of the SCP plans. NRC oversight of the development and implementation of the SCP activities, including a review of DOE's QA program, is designed to help assure that a geologic repository will function as required to protect public health and safety. This oversight is being conducted by the QA and technical staff through ongoing pre-licensing consultation and guidance activites including the On-site Licensing Representive involvement, data reviews and technical meetings.
- 3) Important to Waste Isolation and the Waste Package.
As stated in the 10 CFR 60.151, those buriers important to waste isolation and related activities are subject to QA program requirements specified in 10 CFR 60.152, and therefore comprise a part.of the Q-list. The term " isolation" is i
defined in 10 CFR Part 60 as: " inhibiting the transport of radioactive material so that amounts and concentrations of this material entering the accessible environment will be kept within prescribed limits." Based on this definition and the performance objectives of 10 CFR 60 Subpart E, the term I
" barriers important to waste isolation" (10CFR60.151) means those natural or engineered barriers that contribute to meeting the containment and isolation requirements of 10 CFR Part 60. This rationale and definition was given in the staff's Q-list presentation during the December 1985 QA meeting. DOE Observation 3 and NRC Observation 9 of the meeting minutes were made in response to discussions that followed.
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4 The question raised by DOE was that if they proved that the natural barriers and waste form could meet the containment objectives, would the waste package (we assume they mean those parts which are not the waste form, since the latter is by definition a part of the waste package) have to be on the Q-list. The staff agrees with the State that it is not feasible for DOE to determine with a high degree of confidence what components of the natural and engineered barrier system will meet the containment and isolation objectives prior to conducting site characteri:ation. Therefore all characterization activities which might relate to natural or engineered barriers important to waste isolation should be conducted under a 10 CFR 60.151 QA program in the event that they ultimately are needed to support licensing findings.
- 4) NRC Involvement in DOE Readiness Reviews.
The staff is currently making plans for participation in DOE readiness reviews and, as with other NRC/ DOE interactions, plans to keep the States and Tribes apprised of these plans as they develop. The States' involvement in DOE's readiness reviews should also be discussed with DOE at an early time.
The staff thinks it would be more appropriate for the State to raise this concern with DOE. We note that in a recent correspondence from Vieth to Loux (February 25,1986) that NNWSI has committed to providing you with the QA plans and procedures for the Nevada Project Office and prime contractors of NNWSI.
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e ENCLOSURE 2 Minutes of DOE /NRC Meeting on Quality Assurance December 4 - 5, 1385 8
The meeting was held on december 4-5, 1985 in DOE's Forrestal Building, 1000 Independence Avenue, in Room 1E-245.
Material l
that was discussed at this meeting has been collected and is included as enclosures to these minutes. is an index to the enclosures. is a list of the attendees at the meeting.
The agenda for the meeting is included as enclosure 3.
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- ntroductory re= arks were made by Mr. Sen Rusche, Director of j
DCE's office of Civilian Radioactive Waste Management.
Mr.
Rusche stressed the ccmmitment of DCE manage =ent to quality and noted many recent acccmplishments.
Mr. Jchn Davis, Director of NRC's Office of Nuclear Material Safety and Safeguards, also made introductory remarks.
Mr. Davis reviewed some of the mistakes made by the Nuclear Power industry in quality assurance and encouraged DOE to try to avoid making the same ones.
Mr. Davis also reaffirmed NRC's commitment to provide timely guidance to doe on quality assurance matters.
C DOE Observations 1.
NRC indicated that their involvement with ncn-Q-List items not utilized in the license application veuld be limited to a review and evaluation for the purpose of assuring that none of these items and activities should be on the Q-List.
2.
NRC indicated that the only activities they expect DOE te list in the SCP are major site characterization activities on the Q-List.
Individual tests and experiments would not, in general, be required to be listed.
Major, significant tests, however, will need to be listed in the SCP.
3.
doe felt they might, at some sites, he able to prove that the natural barriers and waste form could meet the NRC containment objective and thus that a wasta package would not need to be on the Q-List.
NRC indicated that the wasta package should be on the Q-List because of the centainment perfer=ance objective in 10 C72 60.ll2 (a) (1).
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CCE requested an opinion from NRC as to whether the DOE
. 1 project office QA organizations met the NRC requirement for independence.
NRC indicated that either of the two arrange-ments DOE projects now have can work.
NRC indicated that they were not yet in a position to determine whether the DOE organization arrangements meet the NRC criterion for QA independence from cost and schedule.
NRC indicated that a variety of different organizational arrangements can work; the key factors are whether the quality message received at the level to which QA reports is as strong as the ccst and schedule messages it receives and the conduciveness of the organizational structure to escalating quality problems co higher levels if sufficient redress is not received at a given level.
NRC's review of the projects' QA plans will address the independence issue on a site-specific basis.
NRC co=mitted*
to reviewing the projects' QA plans as soon as they are submitted to NRC by DOE HQ.
5.
COE presented an overview of.the current size of the DOE project office staffs, the current size of the Q.S. organiza-tiens supporting each project and the projected growth for these.
COE requested feedback from NRC on the suitability of these staffing levels.
No opinion on the adequacy of the numbers was offered by NRC at the meeting.
NRC noted the increase in staffing levels and indicated that a key censideration is the ability of the project to eversee and i
manage the activities and quality assurance prograns of the contractors and participating organizations.
All six of the DOE draft supplements to the OGR QA Plan were 6.
furnished to NRC two weeks prior to the meeting for NRC review and com=ent.
NRC offered general comments during the meeting and committed to provida detailed written com=ents by February 1, 1986.
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NRC was unable to provide to DOE prior to the meeting copies of four NRC Technical Position Papers.which were discussed during the meeting.
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DOE provided to NRC two weeks prior to the meeting a series of questions on Q-List Methodology / Design Guidance and requested NRC's response.
NRC offered general comments during the meeting and NRC committed to provide detailed written comments by February 1, 1986.
9.
DOE is committed to providing NRC a schedule by January 31, 1986, shcWing when NRC can expect:
(1) to receive copies of the ravised CGR-HQ QA Plan and Procedures for review and com=ent.
(3) te receive copies of the ::oE first - rapository pr= ject of: ice QA plans and procedures for review and com=ent.
(3) to receive from DOE the rationale for why the DOE QA programs are considered to be fully qualified and ready for audits.
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i NRC OBSERVATIONS:
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The staff outlined its plans for the next year for giving guidance to COE I
on cuality assurance and assessing the implementation of the CA program.
In order for the NRC staff to not delay the schedules established by DOE for site characterization, 00E should furnish schedules within 60 days for detailed QA program milestones, such as availability of approved 0A plans and procedures for the project offices and prime contractors, plans for additional DOE position pacers or supplements wnich address selected 0A issues, and the DOE rationale that programs are fully qualified and ready for NRC audits.
The NRC staff needs this information for olanning ourcoses so that it may respond quickly to DOE reouests for reviews. This accroach has been previously discussed in the letter from William Purcell, DOE, to R. Browning, NRC, dated September 3,1985, and the NRC's analysis of aerospace technioues acclied to the waste orogram as described in NUREG/CR 2271.
2.
The 00E staff provided responses to most of the issues raised by the NR,C staff during the December, 1984 OA site visits. Several remain to be addressed, however. These issues should be responded to by the DOE in the future, and a schedule for this response orovided. Additional information on these issues can be found in the-meeting minutes for the site visits.
3.
In the DOE letter of November 19, 1985, confirming the arrangements for the December 4-5 meeting on QA, the DOE transmitted nine enclosures related to CA for the repository oroject (See Enciosure 4 to these meeting minutes).
Enclosures 1-6 of the DOE letter are suoplements to the OGR OA Program Plan. Enclosure 7 describes the 00E Systems Engineering Management Plan (SEMP) for the repository project. is DOE's response to issues raised by the NRC staff in the series of site OA visits in December 1984, and Enclosure 9 contains cuestions for the NRC on implementation of Q-list methodology.
During the December 4-5 meeting on 0A, each of these DOE documents was discussed. NRC staff handouts contain bulletized comments on the six supplements to the Or,R OA plan and the SEMP (see Enclosures 14-20 of these minutes). The staff will provide specific written ccmments to 00E on each of the nine Enclosures in the near future (see schedule below). General comments regarding the six supplements are as follcws:
(a) 00E stated that supplements will be developed as the need for them becomes evident. Only two additional supplements are planned.at this time, peer review and cualification of historical data. Drafts of both are to be made available for NRC staf# review in February 19F6 e
N ?q writive sucoierents..md ir revisina tha six ;unciements discussec l
a the meeting, tne COE snould give cafeful consiceration to ensuring t"at the purpose of the supplement is clearly stated, its scooe is clearly defined, and its relationship for-the CGP OA Plan and other d
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sucolements clearly delineated.
The staff noted an impreciseness of language in two of the supplements discussed at the meeting 4
" Calibration of Measuring and Test Equipment" and " Computer Software i
OA" and these steps should both help clarify the language and the i
. intended use of the supplements.
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The staff's schedule for providing written comments to DOE on the enclosures to the DOE letter of November 19 is as follows:
Enclosure Date j
1-6 CA Supplements January 31, 1986 7 Systems Engineering Management Plan (SEMP) January 31, 1986 8 DOE Resconse to Site Visit Issues March 5, 1986 9 0-list Questions January 31, 1986 4
The NRC staff presented briefings on five potential Generic Technical Positions (GTP's) on QA for the repository project. The topics were Configuration Management for Conceptual Designs, Peer Review, 1
Oualification of Existing Data, OA for Research and Exploratory Activities, and 0-list (see Enclosures 8-13 of the meeting minutes).
The staff plans to publish them for public-comment in the Federal Recister in
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early 1986. The staff will accord completion of the draft GTP's ano publication in the Federal Reoister a high priority. For one of the tentative GTP's, CA for Researen and Exploratory Activities, the staff has not reached a conclusion on whether guidance on this topic should be oromulgated in the form of a GTP or some other form. The staff plans, however, to publish for public comment tne other four GTP's.
5.
The subject of audits and cuality program oversight by various levels in the repository program hierarchy was the subject of considerable discussion during the December 4-5 meeting. NRC's experience from the power plant program is that QA audit and management oversight programs often focus largely on paperwork and programmatic issues. NUREG-1055 provides a comprehensive analysis of this problem, its causes, and its results.
In this report, the NRC staff identified comorehensive multidisciplinary team inspections as a particularly useful. tool for the identification of major real or potential cuality or safety problems and fer synthesizing the inputs of technical specialists / inspectors in a
. 1 number of disciplines into a comprehensive picture of the quality of the overall project.
In resoorse to quality, QA, and potential safety problems that developed in power plant design and construction, the NRC developed two headcuarters level team insoection programs, Construction Acoraisal Team (CAT) Inspections and Intecrated Design *nscecticns ('DI'.
Other team ins:ections covering coerstina plants are conducted 'rcr headcuarters !s well. The ICI team inspec-ion accreacn.was described at
- he meeting and COE requested sample copies of IDI team reports.
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3 Accroaches to audits, evaluations, or inscections less comorehensive than IDI's were discussed at the meeting. There was general consensus that, early in a pro,iect, it is important to establish that an adequate QA program, from a programmatic viewcoint, has been established. Once this baseline program has been established, subsecuent audits should focus on implementation of the program.
The NRC staff emphasized the importance of ensuring that audit team membership include representation by people with appropriate technical experience and expertise in the technical areas to be reviewed. The staff also emphasized the need for substantive audits covering technical areas and focusing on program effectiveness, and the iecortance of close attention to, input to, and involvement in audit and evaluation activity by senior management of the organization performing the audit.
The staff referred to several different activities or references that collectively provide cerscective on wnat NRC excects in terns of substantiveness of audits or program reviews and the icentification of root causes of quality and QA problems.
In addition to the IDI's and the data reviews conducted by the NMSS staff, other activities or references.
identified by the staff in this context were the findings of the NRC QA site visits in December 1984 and the QA case studies in NUREG-1055.
Comon threads that run through these evaluatinn methods include the following:
(1) a multidiscalinary team of experts in the major disciplines to be reviewed. The NMSS Division of Waste Manacerent conducts interdisciplinary team data reviews which have similar ab,iectives.
(2) Selection of specific safety systems, activities or QA problems for review.
(3) Comprehensive preparation in the details of what will be reviewed in the field before full field deployment of the team.
(4) Team meetings in which each team members findings and observations are discussed, parallels to other areas are identified, and information is synthesized.
(5)
Involvement of appropriately skilled personnel (e.g., senior management from the reviewing organization) to help aggregate and sort findings, synthesize information, and put results from different disciplines into an overall project perspective.
(5) Communication of the f'indinos, both-in the exit briefing and in we writtec *ecert, to 9igh levels of :"anagecent of the reviewed organi:ation.
0
4 DOE QA managers indicated their intent to perform substantive audits utilizing technical staff in conjunction with OA experts. Several 00E QA managers indicated that some audits of this nature have been conducted already. DOE staff indicated they are develooing a OA auditor course emphasizing the measurement of QA program effectiveness specifically for waste management activities. The NRC staff expressed interest in this course and wishes to be kept informed of progress in its development.
6.
During the meeting, the DOE and NRC staffs discussed the quality assurance information to be submitted or referenced in the Site Characteri:ation Plans. Section 8.6 of the SCP will describe and reference the administrative CA procedures, and Section 8.3 will include and reference information on detailed technical procedures, including the specific implementation of the administrative OA requirements.
The staff believes the general accroach described is acceotaole subject to the follcwing:
the staff is concerned that the traceability of CA recuirements from the administrative procecures to the cetailed technical procedures could be hindered by an insufficient level of detail in the QA administrative procedures referenced in the SCP.
It would be helpful to the staff if examples could be provided before the SCP is submitted showing the hierarchy of documents which define and imolement quality assurance measures. Certain of these documents should also be furnished for staff review.
In a related matter, the DOE and NRC staffs discussed the use of separate OA procedures to accompany the detailed technical procedures, or alternatively, DOE's consolidation of detailed 0A requirements and procedures into the technica! procedures. The staff believes either approach would be acceptable.
7.
In August 1985, the NRC issued Revision 3 to Regulatory Guide 1.28 which endorsed NOA-1 (with minor exceptions) as an acceptable way to meet the OA recuirements of 10CFR50 Appendix B for design and construction of nuclear power plants. The ASME's Committee on Nuclear Quality Assurance (NOA Comittee) has expressed a strong interest in having the NRC staff endorse NOA-1 for application to activities associated with nuclear waste repository.
The DOE through a 00E order has endorsed NOA-1 as describing an acceptable program for meeting DOE QA requirements. During the December 4-5 meeting on OA, the DOE asked the NRC what NRC's plans were for encorsing NOA-1 for waste management, including its schedule.
The NRC plans to endorse with some exceptions, NOA-1 as an acceptable way to meet ecst of the recuirements of Accendix 3 for waste renositories.
The staff does not believe tha-NCA-1 provides sufficient outdanca in some areas certaining to repositories, anc pursuant to 10CFR60 Subcart G, the staff has suoplemented and will further suppiemert the criteria of Appendix B with additional 0A criteria and guidance as applicable.
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5 In the hierarchy of suoplementary guidance on CA, the primary orocram o
reference is and will continue to be the NRC QA Review Plan. Guidance in this plan supplemented as appropriate by staff GTP's and other guidance issued by the NRC staff. The staff plans to endorse NOA-1 for repositories via a GTP.
In the interim, DOE should feel free to use NOA-1 for QA guidance to the extent that it does not conflict with the OA Review Plan and other GTP's and staff guidance that have been or may be issued.
The staffs plans are to publish a draft GTP in the Federal Reaister for public comment in early 1986 endorsing NOA-1 with some exceptions as describing an acceptable program for meeting the OA recuirements of Apoendix 3.
3.
During the meeting, the NRC and DOE staffs discussed the use of readiness reviews fcr assessing the adecuacy of the COE programs, includino the cuality assurance program. The staff believes such reaciness reviews can help to provice a 00E rationale to the NRC that work has oeen or will ce performed in accorcance with NRC regulations.
It has been an apoarently successful technicue employed in non-nuclear applications (e.g.,
aerospace).
NRC oversight of DOE readiness reviews could also provide a part of the basis for the staff's overall evaluation of the DOE ouality assurance program before the start of site characterization. The staff believes such readiness reviews would be an effective and efficient method for the staff to help fulfill its objective of assessing the DOE OA crogram before site characterization. The staff encourages DOE to propose methods for conducting their readiness reviews which would involve NRC staff oversight.
9.
During NRC's presentation on the 0-list, the definition for "imoortant to waste isolation" was provided. The NRC emphasized that the waste package and associated activities are included on the 0-list uncer this definition.
- 10. During discussion of the scope of the Q-list, retrievability was addressed. NRC emphasized that items and activities related to retrievability would need to be considered in develnement of the "0-list."
- 11. NRC notea that in addition to " items", major site characterization activities need to be included in the 0-list as well.
These activities need to be listed to enable the staff to evaluate, in its SCP review, the adequacy of the scope of site characterization activities planned to address the infomation needed to suoport licensing decisions.
- 12. 00E presented issues related to cuestions previously submitted to the NPC on implementation of the 0-14st methodology.
NPC :cmitted to rescordirc to these questions by January 31, 1956.
Followinc ciscussicn of 9ess
.i issuas, both staffs acraei :o the need for a separate meetirc on tha subject. Preliminary discussions icentified two imaortant issues needing follow-up:
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(a) Applicability of a Icw probability cutoff for determining what items
.j and activities are important to safety. COE proposed a frequency of 10 g per year and NRC indicated that this may not be sufficiently e
conservative. The justification for establishing such a limit needs to be exolored further by both DOE and NRC staffs considering such documents as the recent draft ICRP report on waste disposal (ICRP/85/C4-8/12) and NUREG-0612.
(b1 Design basis accidents for developing Q Lists are not explicitly addressed in Part 60. The NRC staff ccrrmitted to evaluating whether Part 50 imolicitly establishes a design basis of 500 mrem or whether the regulation is silent on the issue of design basis accidents and would allcw the NRC flexibility to establish a soecified design basis accice.t. 00E orcoosed a limit of 5 rem, as is curr etly aliowed in Part 72 for determining the controlled area of similar facilities.
13.
In orcer to enable NRC staff to maintain sufficient cognizance of DOE 0A activities and provide guidance in a timely fashion, NRC reouests that it be added to formal distribution of all audit recorts and written responses to same and receive controlled copies of aporoved QA plans and procedures for OGR, OGR project offices, and the prime contractors for each office.
14 DOE presented the methodology recently proposed by Headouarters for grading QA. This methodology includes four cuality levels with grading to be applied within each. Since cuality level one will contain those items and activities on the Q-list and subject to 10CFR60 QA requirements, the NRC staff is interested in the details of applying graded OA within quality level 1.
As noted in the DOE-NRC meeting minutes from the July 1, 1985 meeting on 0-list, DOE is cermitted by Appendix 5 Part 50 to grace OA in accordance with the importance to safety or waste isolation of particular items. NRC also noted that 00E quality levels two through four would also be reviewed but only to assure that the scope of quality level 1 included all items and activities on the Q-list, or to be referenced in or supporting the license application (such as Part 20 recuirements).
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STATE OF NEW MEXICO OBSERVATIONS i
Mr. C. R. McFarland of the State of New Mexico had four comments that he recommends be considered in setting limits for Design Basis Accidents:
2)
Consider the curie content of the Waste Package.
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Consider the fraction of the radioactive material that would be in respirable size particles (i.e., less than about 10 microns) for workers to inhale and for the fence post dose.
3)
Consider the transoort medium and flow path, miticating systems (natural anc engineered), and travel time for enolaced waste - especially where crifts have been backfilled, al Ccnsider K-effective fer worst case analyses.
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$4v Imes E. Ken oy Carl Newton D visicn of Wass n
men' Licensina and Regulatory Division 3
Of ice of Nuclear Mat rial Sa 'ety Office of Geologic Recositories d Safeguards U. S. Department of Energy U
S. Nuclear Regulatory Commis 'on b
W 't & -
wiliarc Altman Division of Quality Assurance, Vender, and Technical Training Center Procrams Office of Inspec*. ion and Enforcemen:
U. S. Nuclear Pegulatory Commission e
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i ENCLOSURE 3 i
1 PRELIMINARY ORAFT 4
NRC "Q-LIST" PCSITICNS FOR PRESENTATICN TO CCE SY S. G. SILh0RN, DECEMBER 5, 1985 1.0 QUALITY ASSURANCE REOUIREMENTS 1.1 "C-List" Items and Activities CCE small acoly sne 10 CFR 60, Suo: art G quality assurance requirements to the items identified as important to safety or waste isolation, and activities relatec therete.
l 1.2 Nen "C-List" Items and Activities For items anc acttvittes wnten are nettner important to safety nor waste isolation but wnich will be referencac in the construction autnerization a ;11 cation to succort fincings recuired by Part 60 (such as recuirements for worner raciological safety ar.d environmental monitoring c:ntainec in 10 CFR 60 Part 20), CCE should describe and reference :ne program for cecumenting and assuring :nat enese requirements have been fulfilled in *
- ne construction authorization aoplication. OCE snould also describe, at leas: in general terms, sucn programs in the SCP.
1.3 Cther Nor "0-list" Items and activities For all c:ner 1: ems anc ac:tytties sucporting the development of a reposi cry, CCE may acoly CA programs based on reliacility, cost, and otner =regrammatic censicerations. The staff will review these non "Q-List" items anc activities only to assure that the "Q-List" is comclete.
1.4 Information Which May Se Used In, or To Succort :ne License Acclication A) CCE snoulc assure tnat all cata collection, interpretation and analyses which may be used in or may support the. license acclication will be performec uncer a QA pr: gram meeting tne recuirements of 10 CFR Part 60, Succart G or, if collected prior to site enaracterization and tne complete implementation of :ne CA program, ce reviewed anc qualified uncer an NRC approved metnoc.
- 3) Prior to and during the early (exploratory) pnases of site characterization wnen the ultimate importance of cata to be collected is not known, 00E should apply a high level of quality assurance to all testing and data collection.
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2-2.0 IDENTIFICATICN OF THE "Q-LIST" CCE shall icentify tne structures, systems, and components important to safety, tne carriers important to waste isolation, anc relatec activities, suen as si:e enaracteri:ation.
2.1 Construction Authorization Acolication The CCE snail icentify a complete "Q-List" in the Construction Authorization Acolication, eitner directly or tnrougn a reference available for staff review.
2.2 Site Characte*ization Plan CCE snoulc icentify a oreliminary list of systems anc major structures and com enents important to safety anc carriers important to waste isolation in the SCP. Major site characteri:ation cata collection activities such as waste package testing, excavation of the exploratory shaft, and surface anc suosurface soil anc rock testing should also ce identified in the SCP, 3.0 METHCCOLOGY FOR CETERMINING EARRIERS IMPCRTANT TO WASTE It0LATICN 3.1 GTP on Licensino Assessment Methocolecy for WLW Geologic Recesitories CCE shoulc use ne performance assessmen: etnocs cescrieec in :ne staff's
" Generic Tecnnical Position on Licensing Assessment Metnocology for nigh Level Waste Geologic Recositories" for determining wnich carriers contricute to or potentially affect the isolation of waste.
3.2 Use performance Allocation Based en Availacle Cata at SCP Stage To tentatively icentify carriers imocr: ant to waste isolation for tne SCP, tne CCE snould allocate performance among the various components of the natural and engineerec systems. Preliminary performance assessments, using tne waste isolation anc containment performance objectives of 10 CFR Part eC and availacle cata, should be utt11:ec *nere practicable as :ne bases for preliminary icentification of carriers important to waste isolation.
4.0 METH000LCGY FOR DETERMINING ITEMS IMPORTANT 70 SAFETY 4.1 Analysis Technioues DOE snoule use the following analysis technicues for determining sne structures, systems, and components important to safety:
Identification of credible events and accident scenarios. Some o
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accicent scenarios might ce so unlikely snat they can be considered
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. increcible, and such accidents need not be considered wnen icentifying items important to safety.
Fault tree / event trees and failure medes anc effects analysis.
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Identification of a source term for racicactive releases anc a rationale for same.
o Accident consequence analysis.
4.2 Cete-mining o- :acilities of Scenarios In cetermining tre aro as111 y of various scenarios, CCE snould use availasle cata for initiating events anc equi ment reliability. Where cata are scarse or unavailacle, bouncing assu=otions should ce usec wits a sue:orting rationale to demonstrate conservatism.
4.3 preliminary Evaluations of Crecible Accider: Scenarios and Their C:nsecuences for :ne SCP In orcer to icentify systems anc major structures and ccm enents in the SCP, CCE should perform preliminary evaluations of credible accident scenarios anc :neir consepuences. Jucgemen: will ce requirec in assessing wnien items are important to safety, and a cr::ablistic aDoroacn may not ce realistic at tnis stage. A senedule for milestones in the cesign acvancement snould ce included in tne SCP.
5.0 GRACED APPLICATICN'CF CUALITY ASSURANCE MEASURES 5.1 Acclication A:penctx 3 of 10 CFR 50, Criterion 2 trcicates tnat the cuality assurance program snall provide control over activities affecttng :ne cuality of ne identifiec structure, systems and components to an extent " consistent witn :neir importance to safety".
5.1 C:rsicerations CCE snoulc apply graded CA measures to items and activities important to safety or waste isolation basec on the follwing considerations:
The impact of malfunction or failure of tne item to safety or waste o
isolation.
Tne c molexity of desig9 or fabrication of an item or the unicueness o
of an item or test.
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The soecial controls anc surveillance needec over processes anc o
equiement.
The cegree to wnien functional como11ance can ce cetonstratec ey o
inspection or test.
1 The quality history and degree of stancardi:ation of the item or o
test.
5.3 CCE may also utilize tne more cetailed guidance on gracing CA measures containec in :ne non-mancatory Appencia 4A-1 of NCA-1.
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L ENCLOSURE 4 DEC 1 2 1985 Dr. Donald H. Alexander Acting Chief Techroloay Branch, RW-23.2 Office of Geologic Repositories V. S. Department of Eneray 1000 Independence Avenue Washington, DC 20585
SUBJECT:
ADDITIONAL NRC COMMENTS ON LEVEL OF DETAIL IN SECTION 8.3 0F THE COE SITE CHARACTERIZATION PLAN
Dear Dr. Alexander:
The U. S. Nuclear Regulatory Commission (NRC) and U. S. Departnent of Energy
/00E) ccnducted a technical meeting on October 29-30, 1985, to discuss section 8.3 of the DOE Site Characterization Plars (SCP). 00E presented their " Content Requirements for Descriptions of Studies in Chapter 8 of the SCP (referred to in this letter as " Content Recuirements") and three examples of study descriptions orepared by each of the three 00E projects using the " Content Recuirements" as a guide. During the meeting NPC provided some prelininary comments on the DOE material docurented in the meetino summary (Enclosure 1) and agreed to provide DOE with additional comments on the appropriate level of detail in section 8.3 and the application of performance coals and confidence levels in the examples. Subsequent to the meeting DOE developed definitions of terms as reonested by NRC in item number 2 of the meeting summary; these were also given to NRC for review in a November 8,1985 letter from D. Alexander to J. Linehan. As acreed to in item number 4 of Enclosure 1, 00E and NRC further discussed during the December 4-5, 1985 meetino on cuality assurance, the quality assurance information to be submitted or referenced in the SCP. NPC comments are documented in the minutes of this meeting.
This letter provides DOE with the results of NRC's review by giving the follcwing additional coments:
Level of Detail in the SCP The comments below conclude that rigorous use of the revised " Content Requirements" (Enclosure 2) will likely result in study plans with the (aq.
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4 IC9/FJ/85/12/09 acpropriate level of detail. Powever, as we have said 'or some time (e.g.,
NUREG-0960 and Regulatory Guide 4.17), it is DOE's decision on the location c' study plans, i.e., if these plans are presented in section P.3 of the SCP or presented as references to section 8.3.
While we agree with 00E's coal of producing a fully integrated and consistent SCP, 00E can best decide how to achieve this while also considerino apornpriate aporoaches to product production. NRC's basic need is to have the plans proviced in some forn 'or review at the time the SCP is issued. This was not done for the SWIP SCR and the Draft EAs (i.e., some key references to these dccuments were supolied much later) and resulted in an inefficient review by the NRC staff.
Soecific Comments on " Content Recuirements" 4
- 1. contains our suggested corrections and deletions to the
" Content Recuirements." These include only some changes which make it more consistent with the Annotated Outline. We suagest that 00E cross check the " Content Recuirements" with Section 8.3 of the Annotated Outline to ensure complete consistency as was agreed to in item number 5 of the meeting summary (Enclosure 11 With the addition of our suggested corrections and additions (Enclosure 2) we consider that the " Content Reouirements" will provide sufficient guidance at this time to prepare substantially comolete drafts of study plans at an appropriate level of detail. We anticipate, however, that there might be details that should be in study plans that are specific to the site, study, test, or analysis. We are available to review this type of material before SCP release, and upon 00E request we will provide feedback and guidance on a case by case basis.
2.
It appears to us that DOE's approach in the " Content Reouirements" assumes 1
that all information on a study (e.g., types, numbers, locations, secuence and duration of tests) should be identified and described in detail and with certainty.
In NUREG-960 and Regulatory Guide 4.17 NRC recognized that plans may be more defined and detailed for more immediate studies and less defined and detailed for more distant studies. We anticipate that for some studies there will be initial uncertainty regardino items such as the location, number, duration of tests or even the most approcriate typa of test. An example of this situation is the hydrologic testing at BWIP.
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109/RJ/85/12/09 An acceptable solution to this case was the develnoment nf a hydrology testing strategy agreeable to both NRC and DOE (9WIP) which is dccumentad in NRC's BWIP Site Technical Position No. 1.1.
This position describes a strategy with decision coints established to determine options for use of various types of tests and various scales and durations of tests. This strategy is particularly well suited for an evolving testing Drogram where there is considerable initial uncertainty.
Frcm the information wa have received it is not clear if other Section 8.3 descriptions (of plans for investigations, specific programs and generic programs) would contain strateoies. Therefore 00E should consider incorporating some form of the testino strategy concect as described above into Section 8.3 at whatever level (s) appropriate. Such an approach would be supplemented by SCP semiannual updates providirq revisions to the strategy and plans as the new information is developed and as decisions are made.
Scecifi; Comments on Project Examoles 1.
We reccgnize that the examples DOE prepared are preliminary and represent i
a first attemDt to apply the " Content Requirements." The evaluations we have developed for each example (Enclosure 3) identify those items in the
" Content Requirements" that are not addressed, addressed to some degree, and which need more information. Considering the preliminary nature of the examples this limited evaluaticn was felt to be appropriate at this time. We believe that the exanples could be made more complete and consistent among projects by revising them based on a consistent and vigorous application of the revised " Content Requirements."
Definitions and Consistent Use of Terms
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We have no additions or corrections to the definitions of terms provided.
However, we believe that the consistency by all 00E projects in the use of these terms and number of hierarchial terms in section 8.3 will minimize.
confusion for those reviewing all three SCP's and will facilitate our review of all three SCP's.
Performance Allocation Although the " Content Requirements" and the project examples relate the information needs to performance goals, none relates the tests to the set e + * *~ -
i 109/:J/85/1^ 09
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oerformance goals and desired confidence levels.
In the September 26-27, 1985 meeting on Subsystem Performance Allocation it was agreed that a rationale would be provided 'or every test or suite of tests in the SCP and that where the tests relate to resolution of performance issues, this rationale would include the relationship of the test to the set performance goals and confidence levels.
(This was noted in the summary of NRC/00E Meating on the SCP's, Section 8.3, Item No. 6.) We recogniza that there has been little tire since the September meeting to incorporate this agreement into SCP cuidance and to implement the guidance. However, this emission makes it difficult to coment in specific terms about wnat level of detail would be appropriate in the examples. We would be cleased to orovide such coments on suitably modified examples. Also, this comment presumes that the references made in the examples to specific perfcrmance goals link to a ecmplete cerformance allocatien in another section of Chapter 8.
Future NRC peviews of Plans It is imcortant to repeat again that for NRC to complete its review of the SCP's in five months, we must at a minimum be current on the existing data for each site. We expect that the extant and nature of our comments on the SCP's will be directly related to the success of our censultations with DOE on plans as they are developed before the issuance of the SCP.
Our preparations for future pre-SCP reviews and early feedback on plans during their development would be enhanced by receiving the section 8.3 hierarchy of plans referred to in the DOE /HO meeting oresentation. This includes the scecific breakdown for each site of generic programs, specific programs, investigations, studies, tests, analyses and procedures.
In addition the Correlation Matrix for each site also referred to in the 00E/HO meeting presentation would be useful early in our reviews to see the inteorated framework of the program expressed by various correlations among tests, 10CFR60, issues, and information needs.
In the DOE letter from W. Purcell to R. Browning (NRC) of September 3,1985, DOE consnitted to meet with NRC in the near future to discuss planned activities, milestones, and appropriate points for consultation with NRC. We consider this step to be critical to planning timely and successful pre-SCP interactions during the next year as well as the post-SCP interactions.
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109/RJ/85/12/09 We believe that the cuality, completeness, ard consistency of the SCP's will be significantly improved by the guidance DOE is developinc in conjunction with NRC's review and comment. While we have noted whera we censider imorovements are needed, the process of DOE /NRC in*eraction on this sub.iect has been appropriate and constructive.
If you have ary questions regardino rur comments please contact R. Johnson at 427-4674 Sincerely,
'UEM !.'".*.'3 TF
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John J. Linehan, Section leader Repository Projects Branch Divisinn of Waste Management Office of Nuclear Material Safety and Safeguards
Enclosures:
1.
Summary of NRC/00E meeting on the SCP's Section 8.3 2.
NRC Mark-up of 00E " Content Reauirements for Descriptions of Studies in Chapter 8 of the SCP."
3.
NRC Evaluatier.s of DOE Project Examples of Study Plans i
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