ML20155E414
| ML20155E414 | |
| Person / Time | |
|---|---|
| Issue date: | 03/07/1986 |
| From: | Linehan J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Knight J ENERGY, DEPT. OF |
| References | |
| REF-WM-1 NUDOCS 8604170597 | |
| Download: ML20155E414 (8) | |
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CF REBrowning MJBell JBunting MRKnapp JTGreeves JLinehan Mr. James Knight RRBoyle Director SCoplan Licensing and Regulatory Division JKennedy Office of Geologic Repositories RCook U. S. Department of Energy Tverma RW-20 PPrestholt Washington, DC 20585 JGiarratana sBilhorn & r/f
Dear Mr. Knight:
Prior to the U. S. Department of Energy /U. S. Nuclear Regulatory Commission (D0E/NRC) Quality Assurance (QA) meeting December 4-5, 1985, the DOE provided the NRC with a series o' 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 D0E. 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 methodology 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 R
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405/SB/86/02/07 are contained in the minutes to that meeting. The GTP is scheduled for publication as a draft document within the next 6 months.
Sincerely, P
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 Record Note: No legal objection per telecon C. Cameron (ELD) to S. Bilhorn, 03/05/86.
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i Response to DOE Questions on Implementation of Q-List Methodology 1.
Allowable Dose Criteria (0perations Phase) 1.1 Question - 60.111(a) and 60.2 indicate that 0.5 rem is the threshold 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(c) 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 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 unrestricted areas at 0.5 rem (10CFR20.105).
10 CFR 60.2 establishes 0.5 rem as the 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.
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 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 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 therefere subject to 10 CFR Part 50, Appendix B.
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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 explicitly establish pre-closure accident dose guidelines. However, 10 CFR Part 60 does specify a dose limit for determining what items will be "important to safety". The term "important to safety" in 10 CFR Part 60 is used to determine what pre-closure items should be on the Q-list. The rationale behind placing a system, structure or component on the Q-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 DOE 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 perform 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 dose 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,-
the staff agrees that the "whole body 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 basis for estimating dose consequences of design basis accidents?
A primary objective at this stage of the repository program should be to I
determine the specific functions and functional requirements of a structure, system, or component anc to identify scenarios which may exceed the functional requirements. The information obtained from tids analysis should then be applied to determine what design details are necessary to assure that the requirements will be met.
As noted in response to Questions 1.1 and 1.2, 10 CFR Part 60 does not 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?
The staff acknowledges that this is a difficult task based on the limited information available upon which to base major decisions. Accident scenarios including initiating events as well as dose consequences for accidents will need to be identified and estimated based on conservative engineering judgment and existing information. The available information base may include data collected and analyzed for other similar activities, such as external events for reactor facilities and design basis accidentr. for ISFSI's and refueling operations at nuclear power plants where these can be shown to apply directly to the HLW facility. AltFough 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.
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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-r 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 latter is a criterion for the underground facility. As noted in the Supplementary Information to the rulemaking establishing the technical criteria in 10 CFR Part 60, the design criterion pertaining to continued operation 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 language could be interpreted to require protection against any fire or explosion that might be physically possible.
48 Federal Register 28194, 28213, June 21, 1983.
Events, internal or external to the HLW facilities, are initiators of accident scenarios.
Internal events, such as eauipment malfunction or operator error, 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.
The term " credible event" would refer to that event which is sufficiently likely to warrant consideration in design of the facility in order to prevent or mitigate the consequences of their occurrence.
2.3 Question - What is an appropriately conservative probability value for credible events / accident scenarios?
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.
The following response addresses a similar question previously posed by DOE:
What does the staff consider an appropriate lower probability limit for accident scenarios considered in the design basis.
It is the staff's position that credible initiating accidents should not be bound by a specific probability 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 the technology, the NRC has factored extremely low probability, high consequence events into their evaluation of the facility.
For example, because of the difference in technology and experience between the Clinch River Breeder Reactor and a typical light water reactor, additional measures were required for Clinch River against accidents beyond the established design basis.
I t' should also be emphasized that in terms of reactor licensing requirements and 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 require 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 appropriate. 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 objectives 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 j
and isolation requirements of 10 CFR Part 60.
10 CFR Part 60 references 40 CFR
6 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),
B.
Components of the natural barrier system,-
C.
Items and activities necessary to support the determination of whether the performance objectives will be met.
D.
Items and activities whose behavior could significantly degrade postclosure performance, and E.
Items and activities important to safety that could affect postclosure performance.
3.2 Question - Should systems, structures, and components important to waste isolati,on 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.
1 3.3 Question - ill the NRC require the application of the single failure criterion to repository facilities prior to closure?
1 As this question relates to preclosure and single failure, it has been addressed in response to Question 2.5 above.
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