Preclosure Safety Analysis-Human Reliability Analysis; Availability of Final Interim Staff Guidance Document, 46682-46688 [E7-16456]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 161 (Tuesday, August 21, 2007)]
[Notices]
[Pages 46682-46688]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-16456]


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NUCLEAR REGULATORY COMMISSION

[HLWRS-ISG-04]


Preclosure Safety Analysis--Human Reliability Analysis; 
Availability of Final Interim Staff Guidance Document

AGENCY: Nuclear Regulatory Commission.

[[Page 46683]]


ACTION: Notice of availability.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is announcing the 
availability of the final interim staff guidance (ISG) document, 
``HLWRS-ISG-04 Preclosure Safety Analysis--Human Reliability 
Analysis,'' and NRC responses to the public comments received on the 
draft document. The ISG clarifies or refines guidance provided in the 
Yucca Mountain Review Plan (YMRP) (NUREG-1804, Revision 2, July 2003). 
The YMRP provides guidance to NRC staff for evaluating a potential 
license application (LA) for a high-level radioactive waste geologic 
repository constructed or operated at Yucca Mountain, Nevada.

ADDRESSES: HLWRS-ISG-04 is available electronically at NRC's Electronic 
Reading Room, at https://www.nrc.gov/reading-rm.html. From this site, a 
member of the public can access NRC's Agencywide Documents Access and 
Management System (ADAMS), which provides text and image files of NRC's 
public documents. The ADAMS accession number for ISG-04 is ML071910213. 
If an individual does not have access to ADAMS, or if there are 
problems in accessing the documents located in ADAMS, contact the NRC 
Public Document Room (PDR) Reference staff at 1-800-397-4209, or (301) 
415-4737, or (by e-mail), at pdr@nrc.gov.
    This document may also be viewed electronically on the public 
computers located at NRC's PDR, Mail Stop: O-1F21, One White Flint 
North, 11555 Rockville Pike, Rockville, MD 20852. The PDR reproduction 
contractor will copy documents, for a fee.
    NRC Responses to Public Comments on HLWRS-ISG-04: In preparing 
final NRC Division of High-Level Waste Repository Safety (HLWRS) ISG 
HLWRS-ISG-04, ``Preclosure Safety Analysis--Human Reliability 
Analysis,'' ADAMS ML071910213, the NRC staff reviewed and considered 34 
comments received from two different organizations during the public 
comment period. Several comments regarded questions about the 
regulatory basis for human reliability analysis (HRA), and perceived 
ambiguity in expectations. Several related comments addressed the use 
of empirical data and their relationship to HRA. The remaining comments 
included recommendations on specific changes to the ISG. Three comments 
on the ISG process were consistent with comments made earlier on HLWRS-
ISG-01, HLWRS-ISG-02, and HLWRS-ISG-03, and were addressed in responses 
to public comment on HLWRS-ISG-01 [see 71 FR 57582, ``Response to 
Comments 13 (a) and (b)''].
    The following discussion indicates how the comments were addressed, 
and the changes, if any, made to ISG-04 as a result of the comments. 
Line numbers in the following comments refer to draft HLWRS-ISG-04, 
ADAMS ML070820387, which was made available for public comment on April 
19, 2007 (72 FR 19729).
    Comment 1. Both commenters noted that ISG lines 79-82 appear to 
imply that ``direct manual operator actions,'' and ``administrative and 
procedural safety controls'' are important to safety (ITS), which is 
inconsistent with the definition of ITS in 10 CFR 63.2. 10 CFR 63.2 
defines ITS as applying to structures, systems, and components (SSCs) 
that are engineered features of the geologic repository operations area 
(GROA), and therefore, actions and controls would not be ITS. One 
commenter recommended specific revisions to ISG lines 79-82.
    Response: NRC agrees with the commenters. However, note that 10 CFR 
63.112(e), which is quoted in the ``Regulatory Bases'' section, in ISG 
lines 196-202, also states that the preclosure safety analysis (PCSA) 
of the GROA must include an analysis of the performance of the ITS 
SSCs. ``This analysis identifies and describes the controls that are 
relied on to limit or prevent potential event sequences or mitigate 
their consequences. This analysis also identifies measures taken to 
ensure the availability of safety systems.'' Therefore, the PCSA 
analyses for ITS SSCs also relate to controls, and measures to ensure 
safety system availability, and these could be tied to human actions.
    The ISG has been revised to change ISG lines 79-82 to: ``Examples 
of human actions that are risk-significant include: (1) Direct manual 
operator actions that are related to reliability of important-to-safety 
(ITS) structures, systems, or components (SSCs); (2) administrative or 
procedure safety controls that are related to reliability of ITS SSCs 
and involve human actions; or (3) human actions that contribute 
significantly to the reliability of ITS SSCs.''
    Comment 2. One commenter stated that there are two broad categories 
of methods to be considered for quantification in HRA: (1) Methods 
based on actual surrogate human performance data from other facilities 
(e.g., chemical processes, interim storage, industrial operations, and 
nuclear power plants); and (2) generic second-generation methods in 
which probability distributions for human reliability are based on a 
qualitative assessment of context and performance factors. The 
commenter states that ISG-04 provides a considerable amount of guidance 
and cautions about the use of nuclear power plant data, but provides no 
explicit guidance on the use of non-nuclear data and no guidance on the 
use of generic second-generation quantification methods [such as 
Cognitive Reliability and Error Analysis Method (CREAM) and Human Error 
Assessment and Reduction Technique (HEART)].
    The commenter recommends adding text to:
    (a) Recognize human reliability data sources in addition to those 
associated with nuclear power plants; specifically, the commenter 
recommends adding a sentence at the beginning of Line 138, as follows: 
``Use of any quantification method, either data-driven or contextual, 
requires justification that it applies to Geologic Repository 
Operations Area (GROA) operations'';
    (b) provide guidance on the use of generic second-generation 
methods.
    Response. (a) NRC agrees with the commenter that data sources and 
approaches other than those associated with nuclear power plants may be 
used as part of the basis for estimating reliability, provided that 
there is sufficient technical justification to do so. The discussion in 
the ISG on the use of nuclear power plant data and approaches, and 
associated justification needed, applies to the use of data and 
approaches from other sources, as well.
    ISG lines 138-139 have been revised as follows: ``Commercial 
nuclear power applications'' has been changed to ``applications for 
commercial nuclear power plants or other facilities.''
    ISG line 142 has been revised as follows: ``Commercial nuclear-
power-plant HRAs'' has been changed to ``HRAs for commercial nuclear-
power plants or other facilities.''
    The sentence in ISG lines 142-145 has been revised to ``Staff 
should expect the use of any quantification method, either data-driven 
or model-driven, to be justified regarding its applicability to GROA 
operations.''
    (b) The U.S. Department of Energy (DOE) has the flexibility to use 
any quantification method it chooses, including CREAM or HEART, 
provided there is sufficient technical basis to use the method for a 
particular application.
    The following sentence is added after ISG line 136: ``DOE has the 
flexibility to choose any method(s) to support the PCSA, given there is 
a sufficient technical basis for applying the method(s) and 
approach(es) to the GROA.''

[[Page 46684]]

    Comment 3. One commenter noted that the term ``risk-significant'' 
is used in many ways in this ISG without a clear definition. The 
commenter recommended the following definition of risk-significant in 
the Glossary: ``Risk-significant: Important contributor to the 
probabilities or the consequences of a single event sequence.''
    Response: NRC agrees with the suggestion to clarify the meaning of 
``risk-significant'' in the Part 63 context, and add a definition for 
``risk-significant'' to the Glossary. Risk-significance would be 
assessed according to those aspects of the LA and technical bases that 
bear on regulatory compliance with 10 CFR Part 63, which is based on: 
(a) Whether an event sequence is category 1, category 2, or beyond 
category 2; and (b) whether the projected consequences meets the 
performance objective for that category. NRC expects the data and 
information provided to be commensurate with supporting these 
determinations. For example, staff expectations will be informed by: 
(1) The extent to which particular SSCs and controls are relied on to 
prevent or reduce the occurrence of event sequences; (2) the severity 
of the potential radiological consequences associated with these event 
sequences; and (3) the potential effects of uncertainty on regulatory 
compliance (e.g., the proximity of the associated frequency to the 
categorization limits for preclosure event sequences, and the proximity 
of the consequences to regulatory performance requirements). See also 
the discussion under ``Uncertainty,'' in HLWRS-ISG-02, p. 4, and the 
discussion under ``Introduction,'' to Appendix A, in HLWRS-ISG-02, p. 
11.
    The ISG has been revised as follows.
    The following has been added to the end of ISG line 64:
    ``The goal of the review is to evaluate whether there is reasonable 
assurance that the performance objectives in Part 63 will be met, which 
in turn is determined by: (a) Whether an identified event sequence is 
category 1, category 2, or beyond category 2, and (b) whether the 
projected consequences meet the performance objective for that 
category. NRC expects the data and information in an LA to be 
commensurate with supporting these determinations, rather than 
supporting precise quantification for all event sequences.''
    The following definition has been added to the Glossary, after ISG 
line 379: ``Risk-significant: Making a significant contribution to the 
probabilities and/or consequences of one or more event sequences that 
have the potential to exceed the performance objectives of Part 63 
during GROA operations.''
    In addition, the following sentence has been added to the beginning 
of the Glossary: ``The definitions provided in this glossary are 
specific to the way the terms are used in this ISG, and may not be 
universally appropriate or applicable.''
    Comment 4. One commenter stated that the term ``full-blown'' is not 
a clear term, and recommended replacing the term with ``full HRA.''
    Response: NRC agrees with the commenter.
    ISG line 56 has been revised to change ``full-blown HRA'' to ``full 
HRA (i.e., encompassing all elements of a complete HRA).''
    Comment 5. One commenter recommended that ISG lines 117-119 be 
revised to delete the phrase, ``Because recoveries are not possible for 
some waste-facility initiators, * * *.'' The commenter stated that the 
reason for reducing the frequency of occurrence of an event sequence or 
minimizing the probability of a hazard is not necessarily because of 
recovery difficulty.
    Response. The intent of the sentence in lines 117-119 is to point 
out that for some waste-facility operational events or initiators 
(e.g., a drop event), recovery actions, such as actuation of safety 
systems to prevent the events-in-progress, may not be possible. 
Therefore, special attention to the associated human-induced initiators 
and the sequence of events leading up to the initiators may be of 
special interest in the staff review of the HRA/PCSA.
    ISG lines 117-119 have been revised to change ``Because * * * 
hazards)'' to ``For waste-facility initiators that may not have 
safeguards to prevent events-in-progress, once initiated (e.g., drop 
events).''
    Comment 6. One commenter stated that the cited nine regulatory 
bases in ISG lines 173-215 do not specifically address HRA within the 
context of the PCSA. The commenter recommended adding the definition of 
an Event Sequence, from 10 CFR 63.2 at ISG line 173, to specifically 
show the regulatory basis for HRA within the context of the PCSA.
    Response: NRC agrees with the commenter's suggestion.
    The following has been added to the beginning of item 1. at ISG 
line 173: ``Event sequence means a series of actions and/or occurrences 
within the natural and engineered components of a geologic repository 
operations area that could potentially lead to exposure of individuals 
to radiation. An event sequence includes one or more initiating events 
and associated combinations of repository system component failures, 
including those produced by the action or inaction of operating 
personnel.''
    Comment 7. One commenter stated that the term ``key'' is used in a 
variety of phrases in lines 220, 223, 446, 461, and 476; yet, the term 
``key'' is not defined and its use in the ISG implies multiple 
definitions. The commenter recommends providing a definition of the 
term ``key,'' in the Glossary, that states, ``Key: Relates to an 
important contributor to the probability or the consequence of a single 
event sequence.''
    Response. The meaning of the term, ``key,'' in the ISG, and 
recommended changes to the YMRP, is the same as it is in plain language 
(i.e., important or fundamental). No further definition is necessary.
    No change was made to ISG as a result of these comments.
    Comment 8. One commenter recommended adding definitions to the 
Glossary for the following terms that are used throughout the ISG, and 
suggested a definition for each of these terms: (a) Human-induced 
initiator, (b) human reliability analysis, (c) pre-initiator human 
failure event, and (d) post-initiator human failure event.
    Response. NRC agrees with the commenter.
    The ISG has been revised, as follows, to add the recommended terms, 
to the Glossary, which begins on ISG line 352, except the term ``Pre-
initiator Human Failure Event,'' which has been defined already in ISG 
lines 375-379 as ``Pre-initiators.'' The ``Pre-initiators'' term has 
been revised to ``Pre-initiator Human Failure Event (HFE).''
    Human-Induced Initiator: An HFE that represents actions that cause 
or lead to an initiating event. The GROA is expected to employ various 
manually controlled waste-handling and transport equipment that may be 
subject to HFEs that could initiate an event sequence.
    Human Reliability Analysis (HRA): HRA evaluates the potential for, 
and mechanisms of, human errors that may affect the safety of the GROA 
operations, including consideration of human reliability as it relates 
to design and programs such as training of personnel. The main 
objectives of the HRA are:
    1. To ensure that human actions that could affect event sequences 
are systematically identified, screened, analyzed, and incorporated 
into the safety analysis in a traceable manner;
    2. Where necessary, to quantify the probabilities of success and 
failure of

[[Page 46685]]

human actions for event-sequence quantification and screening.
    Post-Initiator Human Failure Event (HFE): Post-initiator HFEs 
include both operator actions and inactions that have the result of 
degraded plant/facility conditions. An example of such an HFE is the 
failure to manually actuate or manipulate systems or equipment that are 
required for response to an initiating event, to prevent propagation of 
an event sequence, or to mitigate its consequences. Post-initiator HFEs 
can be further divided into recovery and non-recovery events, as 
appropriate for a given event sequence.
    Comment 9. One commenter stated that the definitions for error of 
commission and error of omission use the term ``degraded plant state,'' 
which does not apply to the GROA. The commenter recommends revising 
lines 354-355 and line 358 by replacing ``degraded plant state'' with 
``event sequence.''
    Response. NRC agrees with the commenter that reference to the 
``degraded plant state'' or ``plant configuration,'' in the definitions 
of Error of Commission and Error of Omission, in ISG lines 353-358, is 
not appropriate for the GROA.
    ISG lines 354-355 and 357-358 have been revised as follows: ``Plant 
configuration'' is changed to ``facility configuration,'' and 
``degraded plant state'' is changed to ``degraded facility state that 
may lead to an event sequence.''
    Comment 10. One commenter stated that although the discussion in 
ISG lines 50-64 is useful, the reference, in footnote 3, to Regulatory 
Guide 1.174, is general in nature and not directly applicable to the 
PCSA. The commenter recommends deleting footnote 3 from the ISG.
    Response. NRC agrees with the commenter that the reference to 
Regulatory Guide 1.174, in ISG footnote 3, is general in nature and not 
directly applicable to the Part 63 PCSA. However, as stated in footnote 
3, the general discussion on the application of NRC's risk-informed 
regulatory principles is useful for other regulatory applications. 
Therefore, NRC disagrees with the commenter's suggestion to delete ISG 
footnote 3.
    No change to the ISG was made as a result of this comment.
    Comment 11. One commenter submitted several closely related 
comments, stating that the draft ISG lacks a sound regulatory basis, in 
that it is built on a presumption that DOE will be conducting an HRA 
that goes beyond what is required by Part 63. The commenter adds that 
the introduction section of the draft ISG on page 1 discusses HRA ``* * 
* as if it were a stand-alone requirement for conducting the PCSA,'' 
when ``* * * HRA should more appropriately be considered one of many 
possible elements of preclosure performance.'' The commenter, although 
recognizing that the paragraph on ISG page 2, lines 50-64, provides a 
more appropriate representation of how the HRA concepts should be used 
by the NRC staff, cites specific examples in the ISG that appear to be 
inconsistent with these concepts. The commenter's examples include the 
use of the phrase, ``the HRA review,'' in ISG page 1; and the mention 
of ``the HRA,'' the ``HRA approach,'' or ``an HRA for the GROA,'' in 
ISG lines 103, 108, 123, 154, and 168, as if a full HRA were required. 
The commenter also states that the ISG statement on page 2 that staff 
should not expect a full HRA including quantification of all human 
error probabilities in the PCSA, seems inconsistent with later ISG 
statements (lines 87-89) which suggest reviewers should verify that the 
HRA for risk-significant processes at the GROA was performed following 
a complete and technically appropriate HRA process, with follow-on 
discussion of ``* * * elements of a highly quantitative HRA process.''
    Response. NRC disagrees with the commenter. Part 63 requires a 
PCSA, supported with adequate technical bases in risk-significant 
areas. Human reliability has been shown to be a key component in 
operations at industrial facilities similar to the GROA. The PCSA 
should address any aspects of human involvement, in pre-closure 
operations, that have a bearing on the performance criteria. The term 
``HRA'' is used broadly to encompass any aspect of the PCSA that 
addresses human involvement. The HRA is not a stand-alone analysis, but 
rather a part of the PCSA that is required (10 CFR 63.112) to 
demonstrate compliance with Part 63 (10 CFR 63.111). As stated in ISG 
lines 2-4 and 50-64, the staff review of human reliability is in the 
context of the PCSA, and is not beyond what is required by Part 63. 
Furthermore, the ISG explicitly states (ISG lines 82-84), ``Staff 
should tailor the scope and emphasis of its review to the approach 
taken in the LA, and the extent to which human actions are (or are not) 
relied on to meet 10 CFR Part 63 performance objectives''; and (ISG 
lines 55-57) that the review should be risk-informed, and staff should 
not expect a full HRA, including quantification of all human-error 
probabilities in the PCSA. Note that even for risk-significant 
processes at the GROA, ISG lines 89-98 explicitly state that the 
quantification HRA steps (c)-(e), may not be needed.
    For clarification, the ISG has been revised as follows.
    The following sentence has been added to the introduction, in line 
7: ``In this ISG, ``the HRA'' refers to any consideration of human 
performance in the PCSA analyses, i.e., the evaluation of the potential 
for, and mechanisms of, human errors that may affect safety of GROA 
operations, including consideration of human reliability, as it relates 
to design and programs such as training of personnel.''
    ISG line 5 has been revised to change ``the HRA review'' to ``in 
the review of HRA in the PCSA.''
    ISG lines 38-39 have been revised to change ``The HRA supporting an 
LA'' to ``The HRA supporting the PCSA in an LA.''
    The following sentence has been added to the paragraph preceding 
ISG line 65: ``Staff should also recognize that the analysis of how 
human performance fits into planned operations and meeting performance 
goals at the GROA may appear in many different parts of the PCSA, and 
in varying scopes (in other words, human performance is likely to be 
addressed in different relevant parts of the PCSA, rather than 
addressed together in one place).''
    The sentence in ISG lines 65-66 has been changed to the following: 
``The guidance in this ISG is written with the expectation that staff 
will seek the assistance of an HRA specialist(s) for review of risk-
significant aspects of an LA affected by human performance.''
    ISG line 78 has been revised to change ``qualitative analyses in 
the HRA'' to ``the qualitative HRA analyses.''
    ISG lines 123-124 have been revised to change ``an HRA for the 
GROA'' to ``HRA in the GROA PCSA.''
    Comment 12. One commenter stated that the ISG imposes, on the 
license applicant (DOE), an expectation that information be provided, 
in the initial LA, that would be more appropriately developed later in 
the licensing and repository development process--and the expectation 
being conveyed by this ISG not only exceeds what is required, but goes 
beyond what is expected to be reasonably available at the time of the 
initial LA. The commenter adds that the programs and processes will be 
developed over time, as the repository moves toward operational status, 
and thus need not be fully developed at the time of the initial LA.
    Response. NRC disagrees with the commenter that the ISG imposes an 
expectation on DOE to provide information beyond what is required to

[[Page 46686]]

demonstrate compliance with Part 63. For NRC staff to review the LA, 
DOE needs to provide sufficient information to demonstrate compliance 
with Part 63--including the basis for safe operations, and where safety 
relies on procedural controls (and human performance), versus hardware 
components. The expectations conveyed in this ISG are consistent with 
10 CFR 63.21(a) that ``The application must be as complete as possible 
in light of information that is reasonably available at the time of 
docketing.'' The Technical Review Guidance contained in the ISG 
provides staff guidance on verifying that appropriate technical bases 
are provided in the LA for the PCSA, with respect to human reliability. 
The subsection, ``Relationship to Programmatic Review and Licensing 
Specifications,'' that begins on ISG line 153, specifically recognizes 
that certain assumptions may need to be verified later and included as 
probable subjects for license conditions in the LA. Note also that Part 
63 requires one LA, with two regulatory decisions: Whether to grant a 
construction authorization in accordance with 10 CFR 63.31, and whether 
to grant the license to receive and possess, in accordance with 10 CFR 
63.41, after construction of the facility is substantially complete. 
NRC recognizes that additional information may become available in 
different stages of the licensing process, but at each stage, DOE must 
provide sufficient information to support that stage. See Commission's 
discussion accompanying issuance of Part 63 (66 FR 55738-9; November 2, 
2001).
    No change to the ISG was made as a result of this comment.
    Comment 13. One commenter submitted two closely related comments, 
stating that the bases, for the parenthetical material in ISG lines 36-
37, and statement in ISG lines 115-116 and footnote 6, discussing 
differences in nuclear power plant versus nuclear materials facility 
operations, are unclear or speculative, since many of the fuel-handling 
operations at the repository will largely be a subset of the types of 
operations carried out at nuclear power plants. The commenter adds that 
there is no reason for NRC to convey additional expectations for HRA at 
the repository over and above what is expected at a power plant and 
suggests that, unless there is a basis, the parenthetical material in 
ISG lines 36-37 should be removed.
    Response. NRC agrees with the commenter that the fuel-handling 
operations at the repository are likely to be similar to the fuel-
handling operations at a nuclear power plant. The intent of the 
parenthetical material in ISG lines 36-37 and the statement in ISG 
lines 115-116 is to compare at-power nuclear power plant power-
generation operations, where rule-based control-room tasks may 
dominate, versus materials-handling activities at nuclear materials 
facilities, where skill-based manual tasks may dominate the operations. 
The reason for these statements is to alert staff to these differences 
since, to date, much of the experience with HRAs, and focus of 
available guidance documents, are on HRAs for nuclear power-generation 
operations (not including fuel-handling activities at nuclear power 
plants). Furthermore, the ISG does not imply staff expectations for HRA 
beyond what is expected for power plants. Therefore, the staff 
disagrees with the commenter's suggestion that the parenthetical 
material in ISG lines 36-37 be removed.
    The ISG, however, has been revised, as follows, to clarify the 
staff's intent:
    In ISG line 36, ``(e.g., nuclear power plant,'' is revised to 
``(e.g., at-power nuclear power-generation operations.''
    In ISG line 37, ``nuclear materials facility'' is revised to 
``nuclear materials facility activities.''
    Comment 14. One commenter stated that the first paragraph beginning 
on ISG line 69 appears internally contradictory, since it first 
discusses the qualitative HRA tasks that are performed as part of an 
overall PCSA (i.e., the conceptual understanding of human performance 
in the planned operations), and then identifies tasks, such as 
identification of HFEs and unsafe actions, as qualitative tasks. The 
commenter stated that: (a) It is not appropriate to describe the 
activities of identification of HFEs and unsafe actions ``* * * as 
qualitative when they are the initial steps of a quantitative 
analysis''; (b) ``Most reliability analysis input for PCSA should not 
require explicit HRA. The reliability of most important to safety (ITS) 
systems, structures, and components (SSCs) should be determined by 
using empirical data collected from similar operations.''
    Response. (a) NRC agrees with the commenter that the tasks 
encompassed by a conceptual understanding of human performance provide 
an important basis of, and hence could be considered a part of, and the 
initial steps, of a quantitative analysis. Similarly, the tasks of 
identification of HFEs, unsafe actions, and factors that influence 
performance are qualitative tasks related to a mechanistic 
understanding of human performance, and can be considered as the 
initial steps of a quantitative analysis. (b) NRC agrees with the 
commenter that reliability analysis inputs based on the use of 
empirical data, in many cases, may not require explicit HRA. However, 
DOE would need to justify that the empirical data are applicable to the 
planned GROA operations, including any human performance aspects. (See 
also response to comment 15 below.)
    For clarification, the ISG has been revised as follows.
    ISG lines 69-74 are replaced with the following sentences: ``It is 
important to have a conceptual understanding of how human performance 
fits into the planned GROA operations and safety. Although quantified 
reliability estimates are typically needed for categorizing event 
sequences, much of the HRA review should focus on the HRA tasks, that 
are performed as part of an overall PCSA, that explain the conceptual 
understanding of human performance in the planned operations. These 
tasks are part of the qualitative HRA analysis and would include, for 
example: (1) Identification of HFEs and unsafe actions; (2) 
identification of important factors influencing human performance; and 
(3) selection of appropriate HRA quantification method(s), if 
considered necessary.''
    The following entry has been added to the ISG glossary: 
``Qualitative HRA Analysis: HRA tasks that include: (1) Identification 
of HFEs and unsafe actions; (2) identification of important factors 
influencing human performance; and (3) selection of appropriate HRA 
quantification method(s), if considered necessary.''
    Comment 15. One commenter submitted several closely related 
comments about the use of empirical data and their relationship to HRA. 
The commenter's statements include:
     A qualitative evaluation justifying the use of empirical 
data for the repository PCSA is a reasonable NRC staff expectation, but 
the applicant should be required to perform quantitative HRA as part of 
the reliability inputs only if human factors were not part of the 
existing data sets.
     ``HRA is only one method of quantifying the human elements 
of risk. A preferable, and likely more accurate, method would be to use 
empirical reliability and event data that quantifies the total 
operational reliability including human influenced circumstances.''
     Regarding the crane data from NUREG-1774, ``Human error is 
implicit in the data. If the applicant can show or commit to programs 
that have comparable rigor to the programs under which the data was 
collected, separate HRA should not be necessary.''

[[Page 46687]]

     ISG lines 428-431 guide NRC staff reviewers to determine 
whether the LA provides justification for data sources, based on 
relevant qualitative considerations--namely HRA activities: (a) 
Identification of HFEs, and associated unsafe actions, to be considered 
in the overall PCSA; and (b) identification of important factors 
influencing human performance. The commenter stated that this is an 
inappropriate implied requirement that is more appropriate for the goal 
of improving human performance, but is not necessary to perform safety 
analysis, and is not required by Part 63 .
     ``Items 4, 5, 6, and 7 on page 14 and 15 are more 
reasonable expectations of the NRC staff review of the repository 
license application than items 2 and 3 (ISG--Appendix A).''
    Response. As noted on ISG lines 50-54, the applicant has 
flexibility in its approach to demonstrate compliance with Part 63 
performance objectives. DOE may choose from a variety of approaches 
that, with adequate technical bases, can successfully demonstrate 
regulatory compliance. Relying on empirical data is one possible 
approach. If the applicant chooses to rely on empirical data to 
estimate reliability of SSCs during GROA operations, staff expects a 
technical-basis discussion to be provided, on why the data apply to the 
GROA operations.
    In addition, see ISG lines 145-150, for guidance on staff review of 
use of empirical failure rates and their technical bases, with regard 
to human performance. If the LA relies on empirical data where human 
performance is an important contributor, the staff expects a 
qualitative evaluation that the relevant conditions at facilities from 
which the empirical data were obtained are similar to those expected at 
the GROA, since HFEs depend greatly on context (see ISG ``Discussion'' 
section, lines 28-39). NRC expects that, as part of this justification 
of ``similarity,'' of the operations at the empirical data facilities 
and the GROA, DOE would include a discussion on conditions relevant to 
human performance, if human performance were an important contributor. 
Item 6, in the Appendix beginning on ISG line 461, also clarifies the 
information pertinent to the data source (NUREG-1774), in the 
hypothetical example and potential discussion, that could be included 
in an LA, to address differences between the GROA and data source 
facilities.
    No change to the ISG was made as a result of this comment.
    Comment 16. One commenter submitted two closely related comments, 
stating that, ``The sections titled, ``Consideration of Applicability 
of Data Approaches,'' beginning on page 4, line 137, and ``Relationship 
to Programmatic Review and Licensing Specifications,'' beginning on 
page 5, line 153, are more reasonable than * * * other parts of the 
Draft ISG,'' with one exception (see next sentence). The last sentence, 
starting at ISG line 167, in the section titled ``Relationship to 
Programmatic Review and Licensing Specifications,'' should be clarified 
or deleted, because: (a) The term ``the HRA'' incorrectly presumes a 
full HRA is necessary; and (b) the phrase ``relevant programmatic 
elements of the HRA'' is not clearly defined.
    Response. (a) See response to Comment 11 above. (b) The purpose of 
this section is to highlight the dependency between HRA and programs 
such as training. Risk-significant elements of the PCSA and HRA that 
rely on assumptions about the adequacy of training and other programs 
are expected to be identified explicitly, and possibly identified as 
probable subjects for license specifications in the LA (requirements 
for future implementation, to ensure that the technical bases of the 
PCSA are valid).
    For further clarification, ISG line 167 has been revised to change 
``programmatic elements of the HRA'' to ``programmatic elements 
supporting the HRA.''
    Comment 17. One commenter was concerned that the proposed additions 
of ``key human actions'' and ``human factors engineering,'' to the 
YMRP, described in ISG lines 220, 223, 292, 296, 300, 304, 308, 312, 
316, and 320, `` * * * may imply that staff LA review should be to 
verify improvement of human performance, rather than to determine if 
regulatory requirements are met.'' The commenter suggested that 
warnings should be placed, in the appropriate sections of the YMRP, 
stating that the purpose of the staff review is to determine regulatory 
compliance.
    Response. NRC disagrees with the commenter. NRC believes that the 
staff understands clearly that the LA review is to verify compliance 
with Part 63, and that the change suggested by the commenter is not 
necessary. The proposed additions of the phrase ``key human actions'' 
to the YMRP are to alert the staff to the need to confirm that 
descriptions of the GROA operations in the LA include the key actions 
that operators would have to perform to maintain safety. Similarly, the 
phrase, ``human factors engineering,'' was added to the YMRP, to alert 
the staff to verify that the quality assurance personnel, assigned by 
DOE to perform independent review of the plans for conduct of normal 
activities, including the written operating procedures, have experience 
and competence in the area of human engineering.
    No change to the ISG was made as a result of this comment.
    Comment 18. One commenter recommended changing lines 252 and 253 to 
(addition italic): ``Verify that any necessary human reliability 
analysis is consistent with * * *.''
    Response. NRC agrees with the commenter. However, the word 
``necessary'' is not included. DOE has the flexibility to choose from a 
variety of approaches for different aspects of the LA. HRA may be 
chosen as one of multiple alternative possible approaches, rather than 
the only necessary approach, for a particular aspect of the LA.
    The ISG has been revised as follows.
    ISG line 108 has been changed from ``the HRA'' to ``any HRA in the 
LA.''
    ISG lines 252-253 have been changed from ``Verify that the human 
reliability analysis is consistent with * * * '' to ``Verify that any 
human reliability analysis in the license application is consistent 
with * * *.''
    Comment 19. One commenter noted that on ISG lines 415-417, the 
concept of ``important to human reliability'' is introduced. (a) The 
commenter stated that this term is not defined in regulation and is 
unnecessary in the draft ISG context. (b) The commenter recommended 
truncating the sentence as follows, `` * * * the data accurately 
reflect the characteristics or features of the GROA,'' in particular 
because it is not just the human reliability aspects that need to be 
accurately reflected in the applicant's safety analysis.
    Response. (a) ``Important to human reliability'' is not introduced 
as a regulatory concept, but rather used as plain language. The purpose 
of this phrase is to remind review staff to keep a risk-informed focus. 
Not every characteristic or feature of the GROA will be important to 
risk contribution from human reliability; the review focus should be on 
those characteristics and features that are significant with respect to 
human reliability. (b) NRC recognizes that human reliability aspects 
are not the totality of a PCSA. The scope of this ISG, though, is 
specifically to provide guidance on reviewing any human reliability 
aspects of a PCSA.
    No change to the ISG was made as a result of this comment.
    Comment 20. One commenter stated that the quote, from NUREG-1774, 
on

[[Page 46688]]

ISG lines 422-427, that the percentage of ``crane issue reports caused 
by poor human performance'' has increased over time and averaged 
between 70-80 percent of the reports, should not be taken to mean that 
human performance is getting worse over time, and NRC should not 
establish any regulatory expectations based on such an assumption.
    Response. NRC is not adding expectations based on these statistics 
reported in NUREG-1774. NRC recognizes there could be many factors, 
known and unknown, that may be driving the statistics. The intent of 
quoting the statistics in the ISG is to show that human performance did 
contribute significantly to the rate of load drops from cranes in the 
empirical data in this hypothetical example.
    To help clarify, the ISG is revised to add the following sentence 
starting in line 425: ``The reason for citing this statistic is not to 
imply that human performance is deteriorating over time, but as an 
indicator that human performance does contribute significantly to 
events in the empirical data in this hypothetical example.''
    Comment 21. One commenter stated that the statement in item 6 on 
ISG lines 491-493 that the NRC staff review should look for a 
``rigorous performance-monitoring program that might compensate for 
elements missing from the NUREG-1774 facilities'' would not be a 
necessary part of the LA unless the applicant claimed better crane 
reliability than the empirical data in NUREG-1774.
    Response. NRC disagrees with the commenter. The ISG does not direct 
NRC staff review to look for `` * * * a rigorous performance-monitoring 
program that might compensate for elements missing from the NUREG-1774 
facilities.'' ISG lines 487-493 discuss a hypothetical scenario where 
there are differences, in the conditions at the facilities from which 
the empirical data were obtained, compared to those at the GROA. The 
ISG lists examples of what the LA might provide as part of the 
technical basis for whatever empirical rate(s) are chosen. ``Rigorous 
performance-monitoring program to account for uncertainties'' is just 
one example of justification the LA may provide for using a particular 
empirical rate (as is, or modified). This is part of the staff review 
of assumptions in the analysis, and checking for justifiable inputs 
from a human performance perspective (which the commenter recognized is 
a reasonable thing to do in the LA review).
    No change to the ISG was made as a result of this comment.

FOR FURTHER INFORMATION CONTACT: Jon Chen, Project Manager, Division of 
High-Level Waste Repository Safety, Office of Nuclear Material Safety 
and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001 [Telephone: (301) 492-3197; fax number: (301) 492-3361; e-
mail: jcc2@nrc.gov]; or Robert K. Johnson, Senior Project Manager, 
Division of High-Level Waste Repository Safety, Office of Nuclear 
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001 [Telephone: (301) 492-3175; fax number: (301) 
492-3361; e-mail: rkj@nrc.gov].

    Dated at Rockville, Maryland, this 10th day of August, 2007.

    For the Nuclear Regulatory Commission.
N. King Stablein,
Chief, Project Management Branch B, Division of High-Level Waste 
Repository Safety, Office of Nuclear Material Safety and Safeguards.
 [FR Doc. E7-16456 Filed 8-20-07; 8:45 am]
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