ML20038C509
| ML20038C509 | |
| Person / Time | |
|---|---|
| Issue date: | 10/13/1981 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| RTR-REGGD-01.033, RTR-REGGD-01.105, RTR-REGGD-1.033, RTR-REGGD-1.105, TASK-MS-901-4, TASK-RE ACRS-1894, NUDOCS 8112110167 | |
| Download: ML20038C509 (12) | |
Text
hWJ-lP9Y DATE ISSUED:
10/13/81 3M s MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON REGULATORY ACTIVITIES
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SEPTEMBER 9, 1981 6; NOV2 019815 3
WASHINGTON, D.C.
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s&'b INTRODUCTIGN j
m The ACRS Subcommittee on Regulatory Activities hcid a meet September 9,1981 at 1717 H Street, N.W., Washington, D.C.
The entire meeting was open to the public attendance.
Mr. Sam Duraiswamy was the Designated Federal Employee for the meeting. A list of documents sub-mitted to the Subcommittee is included in Attachment A.
PURPOSE The purpose of.the meeting was to discuss the following:
1.
Regulatory Guide 1.33, Revision 3, " Quality Assurance Program Requirements (Operation)" (post-comment).
2.
General Revision of 10 CFR 50, Appendices G and H, Fracture Toughness Requirements for Light-Water Nuclear Power Plants (post-comment).
3.
Proposed Regulatory Guide (Task No. MS-901-4), " Identification of Yalves for Inclusion in Inservice Testing Program" (pre-comment).
4.
Proposed Regulatory Guide 1.105, Revision 2, " Instrument Setpoints" (pre-comment).
ATTENDEES ACRS:
C. P. Siess (Subcommittee Chairman), M. W. Carbon, J. J. Ray, H. Etherington.
Principal NRC Speakers:
W. Morrison, W. Anderson, E. Wenzinger, P. Randall, F.. Cherny, T. Scarbrough, S. Richardson, S. Maskeff, J. Page.
EXECUTIVE SESSION Dr. Siess, the Subcommittee Chairman, convened the meeting at 8:45 a.m. and reviewed briefly the schedule for the meeting. He said that the Subcommittee 9112110167 811013 PDR ACRS 1894 PDR
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1 Reg Act Mtg September y
had received neither written comments nor request for time to make oral
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statements on the scheduled items from members of;the public. He infonned the Subco rnittee that proposed Regulatory, Guide 1.13,. Revision 2, " Spent Fuel Storage Facility Desian Basis," oreviously ' scheduled for discussion
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at the subject meeting has been postponed to the next meeting pending re-
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solution of some dissenting technical issues among the NRC Staff.
. 1 REGULATORY GUIDE 1.33, REVISION 3, "0VALITY ASSURANCE PROG 9AM REQUIREMENTS (OPERATION)" (POST COMftENT)
The main objective of Regulatory Guide 1.33,- Revision 3 is to ' describe overall Quality Assurance (QA) program requirements forEthe operational phase of nuclear power plants.
It endorses with certain exceptibns Draft 8 of the ANSI /ANS 3.2-1980 Standard, " Administrative Controls and-Ouality Assurance for the Operational Phase of Nuclear Power' Plants,"-dated Aoril 1981.
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Dr. Siess asked about the status of the ANSI /ANS 3.2 Standard.
Mr. Scarbrough responded that ANS-3 Committeee which is responsible for the development of this Standard is expected to approve it in the very near future; subsequent to the approval of the ANS-3 Committee, this Standard will be submitted to the ANSI Board for administrative approval.
Dr. Siess asked about the appropriateness of endorsing draft Standards in Regulatory Guides.
Mr. Morrison responded that the practice of endorsing draft Standards in Reaulatory Guides is to facilitate the issuance of Reg-ulatory Guides in a timely fashion. There has been some agreement between the ANS Group and the NRC Staff for endorsing draft Standards.
Although they endorse draft Standards in the pre-comment-stage Regulatory Guides, they make sure that a Standard endorsed by a Regulatory Guide is available in its final form prior to issuing that Guide for industry use.
Dr. Siess asked, since this Guide endorses a Standard entitled " Administrative Controis and Quality Assurance for the Operational Phase of Nuclear Power Plants," why wouldn't the NRC Staff use the same title for this Guide instead of using a different title.
Mr. Morrison responded that in the development of the first Regulatory Guide to delineate OA program requirements for the
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s operational:' phase of nuclear-power plantsthe NRC Staff had endorstd two Standar.<h:
ANSI N45.2 for QA program requirements, and ANS 3.2 for admini-stratNEco0trols.' Since ths' dual-Standard endorsement created some con-
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fusion, an. effort was made to combine the QA program provisions of the ANSI N45.2(Standard _into the then ANS 3.2 Standard; consequently, the titles of these kho Standa'rds.had also been combined in one title, namely, "Admini-strative Controls and Q'Jality Assurance for the Operational Phase of Nuclear Power. Pl$nts".
Since,the NRC Staff felt that it would be less confusing to
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use a single title, they have chosen a different title for Regulatory Guide
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1.33, Revision 3 than that of the ANS 3.2 Standard.
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Dr. Siess cemented that using a title different than the Standard that is adorsed by this Guide is not going to help much. Since most people re-member Regulatory Guides by their numbers instead of titles, he does not believe that using a title the same as the Standard is going to create Y...
confusion., He suggested that it would be worthwhile for the NRC Staff to look at the following:
1.
Compile a list of Guides that endorse Standards to see the relationship between.the titles of the Guides and the titles of the Standards.
2.
How many of the existing Guides endorse Standards in whole and how many in part?..If it is partly endorsed, what is the per-centage of the Standard that is endorsed?
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Mr. Scarbrough discussed briefly the changes made to Regulatory Guide 1.33, Revision 3 subsequent to its public comment period. He said that the current verison of this Guide has been revised significantly from the previous version as a result of the resolution of public comments and also to incorporate, as appropriate, the changes made in ANSI' ANS 3.2, dated April 1981, in the follow-ing areas:
1.
Reporting arrangements for QA organizations.
2.
Review and audit programs.
3.
Responsibilities of shift supervision and operating personnel.
4.
Control room access.
5.
Working hour criteria.
Rea Act Mtg September 9,1981 6.
Equipment control.
7.
Emergency procedures format and content.
He said that several Regulatory Positions have been revised to provide clarification, some have been deleted since they have been covered ade-quately in the ANSI /ANS 3.2 Standard, and some new Regulatory Positions have been added to provide additional guidance.
Mr. Scarbrough said that one of the issues still to be resolved is the working hour criteria for the nuclear power plant personnel.
Regulatory Position 8 of the previous version of this Guide endorsed paragraph 5.2.1.6 of ANSI /ANS 3.2 Standard for the working hour criteria. However, pending a Commission decision on SECY 81-440, " Nuclear Power Plant Staff Working Hours," prepared by the Office of Nuclear Reactor Regulation (NRR) of the NRC, Regulatory Position 8 has been revised to delete endorsement of ANSI /ANS 3.2 Standard for working hour criteria. Based on the Commission's decision on SECY 81-440, Regulatory Gide 1.33 will be revised to provide guidance on working hour criteria and the ACRS will be kept informed on the resolution of this issue.
The changes made to various Regulatory Positions and,the reasons thereto are included in Attachment B (pages 1-6).
Dr. Siess asked about the status of the NRC Staff's efforts to clarify the relatioqship between the terms "important to safety" and " safety related".
Mr. iMerison responded that this issue is still under review by NRR.
In the interim period, they are using the definition provided in Appendix A to 10 CFR Part 50 for the term "important to safety".
Indicating that several commentors expressed concern that the six-month implemention period provided in this Guide would be too short to implement l
some of the provisions of this Guide since they would result in comprehensive changes in their QA programs, Dr. Siess asked how realistic is the six month implementation period in view of all the other things the licensees have to do in the next six months or the next year?
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8 Reg Act Mtg September 9,1981 Mr. Scarbrough responded that he believes that many aspects of Regulatory Guide 1.33, Revision 3 can be incorporated by the licensees into their existing programs without much problem.
However, certain aspects would take longer than six months for incorporation and in such cases, the licensees could apply to the NRC for extensions.
Dr. Siess commented that even to write a letter to the NRC asking for extension would require a skilled personnel. He believes that such a skill can be devoted to something else that is probably more important to safety than for appealing for extensions. He asked whether anyone in the NRC is coordinating all of the requirements that are put on licensees by various offices of the NRC and determine how much manpower a licensee has to devote to implement these requirements.
Mr. Silver from NRR responded that there is some coordina-tion being done in this area; however, it is not as extensive as Dr. Siess is alluding to.
He said that, when sending a letter to the licensees with regard to implementation of the provisions of Regulatory Guides, the Division of Licensing of the NRR would take into account to some extent what else that Division has asked the licensees to do.
Dr. Siess commented that the Division of Licensing is not the only office that puts out requirements on licensees. Other offices of the NRC such as I&E are also putting out requirements. A report put out by I&E, " Report on a Survey by Senior NRC Management to Obtain View Points on the Safety Impact of Regulatory Activities from Representative Utilities Operating and Constructing Nuclear Power Plants," dated July 1,1981, points out that some of the factors adverse to safety can be attributed to the enormous number of requirements that have been placed on the licensees, especially in the past two years, with a very short implementation period. Someone in the NRC has to decide which requirements are more important to safety and which should be implemented first.
He believes that there should be some sort of a priority list in implementing all of these requirements.
He asked whether the project manager of a specific plant has a check list of all the things that a particular plant is in the process of doing or has been required to do, ar.d whether he would be able to assess the order of priorities for these requirements. Mr. Silver responded that the Project Manager generally would have a complete list of everything that NRR requires
Reg Act Mtg
-b-September 9, 1981 that plant to do.
He is also aware of the Bulletins and Orders put out by the I&E. However, the Project Manager does not normally assess priorities for these requirements.
It is also highly unlikely that he has a good feel for the total manpower impact for implementing these requirements.
Dr. Siess said that he believes that the ACRS has to keep thinking about this issue. The ACRS may have to bring up the issue of prioritizing implementation schedules either to the Executive Director for Operations or to the Commission in the near future.
After further discussion, the Subcommittee indicated that it would recommend this Guide to the full Committee for concurrence in the Regulatory Positions during the September 10-12, 1981 ACRS meeting.
General Revision of 10 CFR 50, Appendices G and H, Fracture Toughness Require-ments for Light-Water Nuclear Power Plants (post-comment)
Appendix G, " Fracture Toughness Requirements," to 10 CFR 50 specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of Light-Water Reactors (LWRs) to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences.
Appendix H, " Reactor Vessel Material Surveillance Program Requirements,"
includes requirements to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of LWRs resulting from exposure of these materials to neutron irradiation and the thermal environment.
Appendices G and H were issued for industry use in August 1973.
In October 1978, the NRC Staff made some limited revisions to these Appendices to modify the fracture toughness requirements for bolts and to lift restrictions on the location and method of attachment of surveillance capsule holders.
The " general revision" to Appendices G and H is intended to meet se'/9ral long-standing issues.
It updates the requirements of these Appendices to be
s Reg Act Mtg September 9,1981 more consistent with the current technology and pertinent National Standards.
It provides clarification to several requirements of these Appendices and relaxes certain restrictions based on operating e:tperience.
A major part of the revision to Appendix G is in deletion of items now covered in the ASME Code (Section III or IX) and incorporation of the applicable Code provisions by reference.
Similarly, parts of the Appendix H are deleted and replaced by references to ASTM E 185.
Dr. Siess mentioned that the previous version of the general revision to Appendices G and H, which was reviewed by the Regulatory Activities Subcom-mittee on June 13, 1979, as well as the current version, were reviewed by ACRS Consultant Dr. Bush and he did not raise any objections to these revisions.
Mr. Etherington commented that the phrase, "nor lower than the minimum permissible temperature for the inservice system hydrostatic pressure test,"
in paragraph IV A.3 of Appendix G is confusing; in his opinion, this phrase does not impose an additional limit.
It also is not made clear whether the U
NRC Staff's intention is to add 40 F to the limit of 120 F specified in paragraph IV A.2 when the flange condition is controlling. He suggested that additici.31 clarification would be helpful.
Mr. Randall responded that he will consider to Mr. Etherington's suggestion and try to modify paragraph IV A.3, as appropriate, to avoid confusion.
With regard to paragraph IV.B of Apoendix G, Mr. Etherington suggested that a temperature limit of 750 F be specified for thermal annealing treatment of the reactor vessel which he thinks would provide effective measurement of the material toughness properties of reactor vessel beltline.
If a temperature 0
limit of 750 F is not specified, the annealing will normally be performed at 0
650 F (100 F above the normal operating temperature) which is the saturation temperature; but there is little evidence that annealing at this temperature is going to be very effective.
Reg Act Mtg September 9,1981 t
Dr. Siess suggested that it may be appropriate to provide more guidance in the Standard Review Plan for the temperature limit for thermal annealing.
Mr. Etherington commented that the NRC Staff's intent, that 100 percent volumetric examination of the beltline material should also be made, is not made clear in paragraph V.C.1 of Appendix G.
He suggested that some modifications would be helpful.
Mr. Randall stated that he would make approprf ate modifications.
Indicating that in its response to some of the public comments, the NRC Staff speculates that the commentors misunderstood the NRC Staff's intention, Dr. Siess commented that it does not seem appropriate to assume that the commentors misunderstood certain things; he suggested that
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it would be better to contact the commentor to understand his real concern rather than assuming that he misunderstood the NRC Staff's intention.
The Subcommittee indicated that it would recommend the general revision to 10 CFR Part 50, Appendices G and H to the full Committee for concurrence during the September 10-12, 1981 ACRS meeting subject to the incorporation of the changes suggested by the Subcommittee and agreed to by the NRC Staff.
PROPOSED REGULATORY GUIDE (TASK NO. MS-901-4), " IDENTIFICATION OF VALVES FOR INCLUSION IN INSERVICE TESTING PROGRAMS" (PRE COMMENT)
The purpose of this Guide is to provide guidance on the NRC Staff's practice in identifying valves for inclusion in the licensee's Inservice Testing (IST) program and the information needed by the Staff for its review of the program, and on the information needed to evaluate requests for relief from any of the Code provisions.
Mr. Baker said that the development of this Guide was initiated about three years ago in response to a request from the then Division of Operating Reactors.
It does not include any new requirements, but simply formalizes the existing guidclines that have been used by the NRR on a case-by-case basis, i
Reg Act Mtg September 9, 1981 Dr. Siess asked whether the implementation of this Guide would increase the number of valves that have to be tested.
Mr. Baker responded that in some cases the number of valves tested would increase, and in other cases it would decrease.
The main intent of this Guide is to come up with a uniform standard so that the number of valves to be tested by all licensees will be the same.
With regard to the term " safe shutdown" used in the Discussion Section of this Guide, Dr. Siess commented that this term is not clearly defined; it seems that " safe shutdown" as a phrase does not have as clear a meaning as
" hot shutdown" and " cold shutdown" do.
If the NRC Staff's intention is to mean " maintain the reactor safely in a shutdown condition," some modifica-tion is necessary.
Mr. Baker responded that he would make appropriate changes as suggested by Dr. Siess.
Dr. Siess commented that Regulatory Position C.2 and the corresponding paragraph B.2 in the Discussion Section do not relate to each other because Regulatory Postion C.2 refers to valves that perform a pressure isolation function, but paragraph B.2 of the Discussion Section refers to valves that perform both a pressure isolation function and a containment isolation function.
He said that this is confusing and suggested that some modifica-tions would be helpful.
Mr. Baker responded that they will make some changes in paragraph B.2 of the discussion section to avoid confusion.
Dr. Siess commented that the Introduction Section of this Guide gives the impression that this Guide is to implement ASME Section XI requirements for the inservice testing program. However, paragraph B.3 of the Discussion Section and Regulatory Position C.3 seem to imply that this Guide implements also part of Appendix J requirements.
He said that the whole issue is some-what confusing and suggested that the Staff try to make it clear.
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Reg Act Mtg September 9,1981 Mr. Anderson responded that it is not their intent to implement Appendix J requirements in this Guide. They will take a look at this Guide once again and try to clear up the confusion.
Mr. Cherny said that often the NRC Staff gets requests from the licensees for relief from the ASME Seciton XI require-ments, indicating that they will apply Appendix J requirements instead of ASME Section XI requirements; for valves that perform a containment isolation function; the NRC Staff approves such relief requests. Since it is already an accepted practice, it is specified in this Guide.
With regard to paragraph B.5 of the Discussion Section, Dr. Siess commented that the scenarios included in this paragraph are screwhat mixed up and create confusion. He suggested that some effort to make these scenarios clearer is needed. He suggested also that the Staff may reconsider whether they really need to give so many examples for pressure isolation cases in paragraph B.2.
Mr. Etherington and Dr. Siess commented that the sentence in paragraph B.2 of the Discussion Section which states "In almost all such cases, some failure in a redundant system or a failure in the valve itself has occurred," is not clear and does not seem to reflect the NRC Staff's intent. They suggested that some modifications to this sentence would be helpful. The NRC Staff stated that they would make appropriate changes.
With regard to a statement in paragraph B.6 of the Discussion Section which states "A balanced judgement between the hardship and compensating increase in the level of safety must be explicitly justified if, for example, an extended shutdown period is required to test all the valves at the prescribed code interval," Dr. Siess commented that it is not clear who has to do the balancing.
It also seems to imply that the judgement of the NRC Staff has to be justified, which he believes, is not the intent here. He suggest'ed that the Staff rrodify the language to reflect their real intent.
Regarding Regulatory Position C.1, Dr. Siess commented that the use of two different words "needed" and " required" is unnecessary.
Since everything that is required by the NRC is presuaed to be needed, he does not believe that they need to use the word "needed".
Further, the use of the phrase
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g Reg Act Mtg September 9,1981 "the use of voluntary standard" in C.1 does not seem appropriate. He suggested that the NRC Staff consider using some other phrase or just delete this phrase.
The Subcommittee suggested several other editorial changes, provided guidance for clarification and improvement in certain sections of this Guide and indicated that the NRC Staff could issue this Guide for public comment.
PROPOSED REGULATORY GUIDE 1.105, REVISION 2, " INSTRUMENT SETPOINTS" (PRE COMMENT)
The main objective of this Guide is to describe a method for ensuring that instrument setpoints in systems important to safety are initially within and remain within the specified limits.
It endorses, with very limited exceptions, the Instrument Society of America Draft Standard ISA-ds67.04, entitled "Setpoints for Nuclear Safety-Related Instrumenta-tion Used in Nuclear Power Plants".
Dr. Siess asked about the status of the Standard ISA-ds67.04.
Mr. Wenzinger responded that it has been submitted to the Standards and Practice Department of the ISA for approval. He said that the NRC has received a letter from the ISA authorizing them to endorse the draft ISA-ds67.04 Standard in Regulatory Guide 1.105, Revision 2; a copy of the ISA letter is included in Attachment C.
Dr. Siess suggested that it might be helpful if the NRC Staff could bring an instrument and explain to the Subcommittee why and how the setpoint drift occurs so frequently.
Mr. Wenzinger said that setpoint drift can be caused by a number of mechanical as well as electrical problems.
It can be caused by:
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a change in calibration of a sensor, 2.
a change in the value of a power supply output voltage, 3.
environmental conditions that might affect the current that is produced by the transistor, and 4.
aging of a resistor.
r Reg Act Mtg September 9,1981 Mr. Wenzinger pointed out that Revision 1 to Regulatory Guide 1.105 did not address the question of providing an adequate margin between a nominal trip setpoint and the allowable value. Revision 2 to Regulatory Guide 1.105 and ISA-ds67.04 Standard address this issue specifically.
Af ter further discussion, the Subcommittee indicated that the NRC Staff could issue this Guide for public comments.
FUTURE MEETING The next Regulatory Activities Subcommittee meeting is scheduled to be held on October 14, 1981 to discuss the following:
1.
Regulatory Guide 1.23, Revision 1, " Meterological Programs in Support of Nuclear Power Plants" (post comment).
2.
Proposed Amendment to 10 CFR Part 50, Section 50.55a, " Codes and Standards" (pre comment).
3.
Regulatory Guide 1.13, Revision 2, " Spent Fuel Storage Facility Design Basis" (pre comment).
4.
Proposed Regulatory Guide (Task 1C 121-5), " Response Time Testing of Protection System Instrument Channels" (pre comment).
5.
Proposal by Mr. Bender for an cbbreviated Safety Analysis Report.
Dr. Siess thanked all participants and adjourned the meeting at 12:38 p.m.
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NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room,1717 4 Street, N.W.,
Washington, D.C., or can he purchased from Alderson Reporti',9 Company, j
Inc., 400 Virginia Avenue, S.W., Washington, D.C. 20024, (202) 554-2345, t
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