ML20056D440
| ML20056D440 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/21/1993 |
| From: | Mcgaha J ENTERGY OPERATIONS, INC. |
| To: | Gillespie F NRC |
| References | |
| FRN-58FR29012, FRN-58FR33285 NUDOCS 9308160214 | |
| Download: ML20056D440 (11) | |
Text
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" ENTERGY
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John R. McGaha July 21,1993 Regulatory Review Group U.S. Nuclear Regulatory Commission Washington, D.C. 20555
.g ttention:
Mr. Frank P. Gillespie, OWFN 12 D21 A
Subject:
Comments on NRC Regulatory Review Group Draft Report CNRO - 93/00026
Dear Mr. Gillespie:
E!ntergy Operations, Inc. has reviewed the NRC Regulatory Review Group (RRG) draft report concerning the review of power reactor regulations and related issues (58 FR 29012 and 58 FR 33285). We wish to submit the following on behalf of Arkansas Nuclear One Units 1 & 2, Grand Gulf Nuclear Station, and Waterford 3 Steam Electric Station.
Overal', the report is of a high quality and contains a number of excellent recommendations. The NRC management and staff deserve recognition for such an in-depth review and the resulting innovative alternatives to current regulatory approaches. We believe that this initiative, if continued to be pursued, will ultimately result in a significant benefit to safety for the commercial nuclear power industry.
While difficult to quantify, the current level of regulatory complexity and burden, coupled with overcommitmer.t by licensees, represents a source of distractions to both our personnel and our allocation of resources at the expense of more safety significant concerns. Removing unnecessary requirements and streamlining regulatory processes will result in allowing the industry and the NRC to more clearly focus on operational safety issues.
We urge the NRC to pursue these and additional efforts related to regulatory burden reduction. Such efforts are extremely important to the future of the commercial nuclear power industry. Because of the limited lifetime of the RRG, organizational b
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i Comments on NRC Regulatory Review Group Draft Report i
July 22,1993 i
CNRO-93/00026 Page 2 of 2 i
and policy changes must be expeditiously implemented by the NRC to ensure that the work of the RRG continues. We also believe that in order for the RRG effort to continue a continuing charter and champion within the NRC will be required.
in addition to NRC efforts, we recognize that many changes recommended by the 1
RRG will require licensee support and participation. Entergy Operations is ready to provide needed support in the further development of recommended initiatives and f
would appreciate the opportunity to participate with the staff in this effort.
I Additional comments are provided in the attachment. We appreciate this opportunity to express our views on the report and the Commission's consideration of our comments. Please contact Mr. Kenneth Hughey (601-984-9756) or Mr. Herbert Kook (601-984-9766) of my staff should you have any questions or desire additional I
information regarding this matter.
Sincerely, S.r gR, McGALA JRM/wkh attachment cc:
Mr. T. W. Alexion Mr. J. L. Milhoan Mr. R. P. Barkhurst Mr. P. W. O'Connor Mr. R. H. Bernhard Mr. N. S. Reynolds Mr. R. B. Pevan, Jr.
Ms. L. J. Smith Mr. J. L. Bicunt Mr. D. L. Wigginton Mr. S. D. Eb:teter Mr. J. W. Yelverton Mr. E. J. Ford Central File (GGNS)
Mr. C. R. Hutchinson DCC (ANO)
Mr. H. W. Keiser Records Center (WF3)
Mr. R. B. McGehee Corporate File [ 12 )
Comments on NRC Regulatory Review Group Draft Report July 22,1993 Attachment to CNRO-93/00026 Page 1 of 9 Additional Entergy Operations, Inc. Comments On the NRC Regulatory Review Group Draft Report yolume One Page 5, para 2 The statement that " full implementation of an attemative approach" to Appendix B will not require a rule or license change is not necessarily true for all possible alternatives to the existing approach, although such a statement does appear to be true for those approaches discussed in the report.
Page 6, section 4.1.1 Care should be exercised with this and other sections so that new definitions do not lead to a new set of requirements which are more burdensome than the existing requirements. New definitions should be carefully made and fully coordinated among NRR, the NRC Regions, and the industry before their adoption.
Page 7, first bullet it may be premature to state that Appendix B itself should not be changed or modified. More review and consideration, and discussion with the industry, should occur before this is stated definitively.
Page 7, second bullet Clarify that the Regulatory Guides 1.84,1.85, and 1.147 are those which appear to require revisions on a periodic basis.
Page 10, first bullet Clarification should be added to the phrase " maintaining appropriate control of changes to material that is removed from Technical Specifications.. " EOl assumes this refers to Core Operating Limit Reports, certain surveillance requirements, equipment lists, etc. placed under licensee administrative controls, but a more explicit description is needed.
Pages 17-19, section 4.3 The ideas in this section represent a significant potential for reductions in unnecessary burdens for the industry. However, the industry and the NRC have many obstacles to overcome before the results outlined in this section are achieved. A recommendation should be considered for the NRC to evaluate offering introductory training in risk technology for the staff.
The fourth bullet on page 18 recognizes a key obstacle to risk technology use by the industry, i.e., the NRC embracing risk technology where it shows additional requirements are necessary but rejecting it when it contradicts a
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. Comments on NRC Regulatory Review Group Draft Report July 22,1993
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Attachment to CNRO-93/00026 Page 2 of 9 preferred staff position. A Policy Statement, SRM,'or other process might be used to improve this situation. For example, the staff might be required to i
specifically address risk considerations in the documented backfit evaluation or analysis for new requirements.
j Volume Two t
Pages 9-10 The discussion of commitments and changes to commitments should be clarified. The term " equivalence in safety" and " equivalent to the original commitment" should be replaced with discussion such that commitment changes are permitted without NRC approval so long as the changed commitment does not constitute an unreviewed safety question per 10 CFR 50.59.
Page 10 in the first sentence, " Proposed " should be deleted (also in the third sentence) and "shall be" changed to "are" since 10 CFR 50.71(e) already requires the reporting of changes discussed but does not require reporting prior to the change being made. In section 11, the use of" prospectively"is unclear; if it refers to applying the new change method and definitions to future G
commitments only, the NRC, industry, and licensee use of multiple commitment definitions and processes may be extremely cumbersome.
Pages 25-26, sections IV and VNI Renumber the second section V to VI and the existing VI to Vll. While considering changes to the NRC FFD program, the NRC rejected a two-tier system of 50% testing for employees and 100% for contractors for a single system at a 50% rate. Other industries regulated by the DOT have a significantly higher positive test rate than nuclear employees or contractors; the DOT considers these rates adequate to propose to lower their required testing rate below 50%. Therefore, the recommendations in section VNI should be changed to include that the rulemaking discussed in section IV should be changed (or a new rulemaking proposed) to allow 50%
testing for all nuclear workers.
j Pages 25-26, sections V-VNI Overall, the proposed recommendations appear positive. However, the NRC RRG should also consider the burden caused by
.l the scope of audits, auditing HHS-approved laboratories, and/or the re-auditing j
of activities and information pertinent only to laboratory selection and establishing initial contracts with laboratories.
Comments on NRC Regulatory Review Group Draft Report July 22,1993 Attachment to CNRO-93/00026 Page 3 of 9 Page 26, section VINil There is no basis for the recommendation provided under Potential improvements and the need for this action is not evident. In addition, this appears to recommend establishing a new regulatory requirement via generic communications and contrary to 10 CFR 50.109. New regulatory requirements should be established through rulemaking and in compliance with 10 CFR 50.109 as discussed elsewhere in the report.
Pages 27-30 Generic communications are inconsistently used, which leads to NRC and industry confusion. Additional action is needed to resolve prior industry concerns in this area. New definitions should be adopted by the NRC for the types of generic communications, for example:
-Information Notice: A mechanism to share NRC and/or industry experience issued for information only; new/different NRC positions or new regulatory requirements would not be established by Information Notices
-Bulletin: A communication of action (s) in regard to a specific issue which the NRC would clearly expect licensees to either complete or provide justification for alternative action (s); these expectations may involve a new or different NRC position regarding previously existin 1 regulations but new regulatory requirements would not be imposed by a Bulletin; confirmation of actions or siterr.stives may be requested under 10 CFR 50.54(f) (note much of this role has been recently performed by generic letters)
-Administrative Letter: A communication of an administrative or informational nature which would not require a response and would not address: new/different NRC positions, new regulatory requirements, or generic safety issues
-Generic Letter: A communication about generic safety issues, or a generic request for information which would be reasonably expected to exist (possibly under 10 CFR 50.54(f)), and which might involve new/different NRC positions. Generic letters would not: establish new regulatory requiremems, request confirmation of the completion of new actions / programs, or require the generation of significant amounts of new information.
Page 29 We believe that the NRC staff biweekly letter listing pending generic communications could be changed to a monthly letter without impact on the industry while possibly reducing NRC burden.
Comments on NRC Regulatory Review Group Draft Report July 21,1993 Attachment to CNRO-93/00026 Page 4 of 9 Pages 64-65, section 2.3.8 The regulatory basis for " requiring" the submittal and review by the NRC of the computer code verifications is unclear. The only basis mentioned is Generic Letter 83-11, which is not a regulatory requirement.
information should be provided for the basis of this requirement, the i
recommendation changed so that the NRC staff no longer imposes this burden upon the industry, or rulemaking initiated to properly establish the requirement.
1 l
Pages 68-69, section IV The text of section IV states that the rules goveming i
l changes to QA, security, fire, and emergency plans should be the same. The changes in approved QA programs to be submitted for prior NRC approvalin the proposed 10 CFR 50.54(a)[3] do not appear consistent with the changes i
requiring prior NRC approval recommended for 50.54(p)(2) and 50.54(q).
Change "provided the change does not reduce the commitments in the program description previously accepted by the NRC" to "provided the change l
does not reduce the commitments in the program below the requirements of 10 CFR Appendix B."
i l
Page 71 section Ill Consistent with a philosophy of eliminating unnecessary regulation and regulating only where real need exists, this section should not be included unless real situations occur in which a utility fails to provide adequate space for NRC personnel.
Pages 73-74 See comments regarding pages 9-10 and page 10.
Pages 75-81 As inferred in this section, Policy Statements addressing licensee issues are frequently considered by the NRC staff as regulatory requirements.
Such requirements should properly be imposed upon licensees as backfits in accordance with 10 CFR 50.109. Add that Policy Statements should address only internal NRC policies, i.e., practices and expectations, rather than those expected or demanded of licensees. Items in section A should be deleted, and items in sections C, D, and E should be deleted or replaced by rulemaking.
1 Pages 90 - 92, section ll These are strong statements, which we endorse. They do, however, highlight the need for the NRC to be willing to adopt changes in attitude and culture to accept the new approaches discussed in the report.
Page 98, section Xil in this section and many others, the term " performance-based" is frequently used but is never well defined or explained. As discussed previously, various people or organizations in the NRC and the industry may
Comments on NRC Regulatory Review Group Draft Report July 21,1993 Attachment to CNRO-93/00026 :
Page 5 of 9 make a number of different interpretations of any term which is not well defined. The report should include a clear definition of performance-based which is agreeable to the industry and the NRC.
Page 107 The reference for " update of the FSAR" should be 10 CFR 50.71{e)(4).
Page 145-146 The discussion on this subject is excellent. The recommendation y^
should be modified to encourage licensees to begin considering the use of risk e
methods to provide additional bases for USQ determinations; this would allow the NRC and licensees to gain experience and insight in using these tools as soon as possible; in addition, some concerns not apparent by more traditional methods might be identified.
Volume Three No comments.
Volume Four General There is a considerable amount of new and different materialin this section.
t it should be carefully considered and reviewed before pursuing full implementation. And as mentioned previously, care should be exercised so that new definitions do not lead to a new set of requirements which are more burdensome than the existing requirements. New definitions, especially in this relatively new area, should be carefully made and fully coordinated among NRR, the NRC Regions, and the industry before their adoption.
{
Section 4.2.3 Human reliability analysis (HRA) offers valuable insights to a PRA study. By j
using HRA analysis, past PRAs have identified a number of problems in procedures and training that had not been found using " traditional regulatory" j
approaches. HRA techniques can help to measure many effects which influence the reliability of an action, thereby establishing the relative j
importance of human failure events. This assists _in prioritizing training 1
i activities and showing the effectiveness of verification activities, allowing the most effective use of resources. HRA techniques offer the best, if not only, means of evaluating the significance of training, stress, hesitancy, conflicting priorities, etc.
Comments on NRC Regulatory Review Group Draft Report July 21,1993' Attachment to CNRO-93/00026 Page 6 of 9 If an Human Error Probability (HEP) of 1.0 is used, the effectiveness of using safety prioritizations determined by the IPE is weakened. Most application work and sensitivity work done (by utilities) with PRA is done by manipulating the cut set results. Therefore, many cut sets are lost (i.e., truncated) once an HEP has been added to the cut set. Evaluations donc by the manipulation of dominant cut sets usually can be performed fairly quickly. However, requantification of an entire PRA is much more time consuming.
if the PRA should be requantified with all HEos (both pre-accident and post-accident) set to 1.0, this may well cause impodant cut sets to be lost because of limitations of the computers used in quantifym2 the accident sequences.
That is, important hardware cut sets may be truncated because of the large number of cut sets generated because of the overly conservative HEP of 1.0.
Rather than setting all HEPs to 1.0, a reasonable screening value should be used. Different screening values should be used for pre-accident and post-accident HEPs. The pre-accident screening value could be relatively low as these actions are usually controlled by procedure and well understood. Post-accident screening values of approximately 0.5 could be used since these are usually cut set dependent and time considerations come into play. Values or ranges of values should be specified by the NRC or the industry for I
acceptance without specific additional plant-specific justification.
Screening values are often used in PRA studies to solve the plant model in order to protect against dependencies, high stress, and other influences which could invalidate a human failure event probability. An a iequate consensus should exist resulting from this experience to develop industry screening values. The NRC Accident Sequence Precursor study method values for human failure events might also be used; these are regularly used by the NRC staff in regulatory evaluations. Another option would be the development of consentative probabilities developed for actions required within various time intervals. Such an approach would allow the distir;ction between actions required to be performed in ten minutes from those required to be performed in several hours.
Sectior 4.4 Consideration of a graded approach to QA programs based on risk is a very positive development in allowing the industry to utilize resources commensurate with importance to safety. However, the benefits to be gained
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Comments on NRC Regulatory Review Group Draft Report July 21,1993 Attachment to CNRO-93/00026 Page 7 of 9
. are a function of how the QA groups are defined. The proposed groupings throughout section 4.4 appear overly conservative and will limit the benefits from adopting such an approach. Specifically, including any SSC found in "A PRA," that is, in any similar plant, in the most important QA group will unnecessarily reduce the benefits of a graded program and will conflict with PRA application to the Maintenance Rule.
Although the NSSS design of plants are similar to other plants of their class, there are significant design differences, garticularly in support and balance-of-plant systems, such that the validity of applying the conclusions of one plant PRA to another is doubtful. For example, a support system at Grand Gulf may exhibit a significantly different risk measure than the same system at another i
BWR/6; requiring Grand Gulf SSCs to be determined by the characteristics of another plant is inappropriate. An attempt to differentiate these SSCs is made in the document but results in the creation of another category of SSCs. SSCs that are shown to be not important on the basis of a plant specific IPE do not warrant being included in a graded QA category higher than SSCs not important in any PRA.
in addition to the above concern, some plants such as Grand Gulf have utilized their IPE results to rank SSCs for the Maintenance Rule implementation.
Adopting a different approach to a graded QA program would create inconsistencies between these two programs, while they should have the same basis.
i If the goal of including risk results from other plants is to ensure that all necessary SSCs have been included with a minimum of review effort, the same result could be attainable through a licensee comparison of plant-specific risk rankings against " class-average" rankings and requiring justification for any discrepancies which might be of concem. Most plants have already compared their IPE results with other facilities of their class and understand the reasons for such differences. Therefore, little additional burden would be created. This appears to be a much more preferable approach than mandating unnecessary conservatisms which would be in effect for the remaining life of a plant.
Section 4.4.1 i
The need to be able to normalize the importance measure is understandable but this appears to put plants with better core damage frequency numbers at a disadvantage to those with poorer results. For example:
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Comments on NRC Regulatory Review Group Draft Report July 21,1993 Attachment to CNRO-93/00026 Page 8 of 9 Plant #1 has baseline CDF of 1E-5 with an SSC that has an Achievement importance Measure of 4E-4. The unavailability impact for this SSC is a factor of 40 which would be considered relatively important for Plant #1.
Plant #2 has baseline CDF of 1E-6 with an SSC that has an Achievement importance Measure of 4E-5. The unavailability impact for this SSC is again a factor of 40 which would also be considered relatively important per the definition.
However, the change in CDF associated with Plant #1's SSC (Achievement importance Measure - CDF) is 3.9E-4, while the change in CDF associated with Plant #2's SSC is 3.9E-5. There is an order of magnitude difference in the two but the defined Achievement importance Measure impact ratio makes them appear equal. There should be another measure to capture importance without penalizing those plants that have a relatively low CDF. Possibly, the above could be offered as an example with some desirable attributes, such as providing a method to ensure that existing levels of safety are maintained.
Section 4.4.2 - 4.4.3 On page 4-27, second para The first two sentences would be clearer if changed to: "Those SSCs of a plant which are not included in the Q list have been determined to be deterministically unimportant, and are therefore not currently required to be subject to QA requirements."
As discussed above under Section 4.4, the inclusion of all SSCs found to be important in "a PRA" is not necessary if justified. Also, as now written, the description of Group A SSCs appears to envelop all Group B SSCs, which appears confusing. A possible alternative phrasing (which also addresses the preceding comment) might be:
" Group A - Those SSCs meeting one of the following criteria:
-Found to be important in the plant-specific PRA
-Found to be important through deterministic methods.
Group B - Those SSCs meeting one of the following criteria:
-Found to be important in a PRA of a plant of similar design and which has not been shown to be not important by plant-specific means
-Found to be important to plant or system availability through deterministic or other means."
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4 Comments on NRC Regulatory Review Group Draft Report July 21,1993 Attachment to CNRO-93/00026 Page 9 of 9 For Group A SSCs, generally the same standards as now are used for SSCs requiring " full QA" would be used for design, procurement, receipt, storage, installation, testing, maintenance / surveillance, etc. For Group B, only limited activities such as design control and initial testing should be imposed. Other programs and requirements (i.e., the Maintenance Rule) would ensure that these SSCs would continue to operate to the level identified as appropriate by the IPE. For Group C, these would be subject to the same minimal requirements as now applied to "nonsafety-related" or nonnuclear equipment.
Section 4.4.4 No basis is provided for the statement that " Generally, updating the PRA at every refueling outage will provide this [necessary) currency." As the industry matures, significant changes to existing nuclear power plants are expected to be greatly reduced if not eliminated. Evaluating changes as they occur, as part of the design and licensing process, for possible impact on a PRA would appear to be a more reliable and less burdensome process than specifying an arbitrary period for updating a PRA. Depending on the changes and PRA use at a given plant, there may be occasions when updating once an outage is not frequent enough or occasions when updating is not needed for a number of outages. At this point in the development of risk technology, such a statement appears to be contrary to many of the industry and RRG concerns abo'2t overly prescriptive and arbitrary regulatory requirements.
Section 4.5.2 Although the concept of using CDF to increase STis for relatively non-important SSCs appears sound, the assumption of unavailability as directly proportional to time x failure rate may be f! awed. Some component failure rates, such as those for diesel generators and normally nonenergized solenoid valves, may be more valid when computed per number of demands rather than per time period. Also, the equation used does not address unavailability due to out-of-service time for maintenance and surveillance activities, which might be greater in some cases than that due to failure rate x STI. For example, the equation seems to imply that reducing an STI by one-half would reduce unavailability one-half; there does not seem to be any consideration of a
" break-even" point for continuing to reducing the STis of important SSCs.
Section 4.5.3 See previous comments on Section 4.4.4 and 4.5.2.
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