ML20247Q914
| ML20247Q914 | |
| Person / Time | |
|---|---|
| Issue date: | 05/31/1989 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| Shared Package | |
| ML20247Q900 | List: |
| References | |
| ACRS-GENERAL, NUDOCS 8906070081 | |
| Download: ML20247Q914 (36) | |
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ACRS COMMITTEE AND CONSULTANT REPORTS - 165 Submitted in Accordance with Sections 10 and 13 of the Federal Advisory Comittee Act May 1989 8906070081 890601 l
b C0NTENTS DATE I).
SUMMARY
LETTER 05/26/89 II) ACRS REPORTS TO CHAIRMAN, NRC 1.
Remick ltr to Zech, NRC's Human Factors 05/09/89 Programs and Initiatives 2.
Remick ltr to Zech, Generic Letter on Safety-05/09/89 Related Motor-Operated Valve Testing and Surveillance 3.
Remick ltr to Zech, Generic Letter Related to 05/09/89 Occupational Radiation Exposure of Skin from
- iot Particles 4.
Remick ltr to Zech, NUREG-1150, " Severe Accident 05/09/89 Risks: An Assessment for Five U.S. Nuclear Power Plants" 5.
Remick ltr to Zech, Operating License Application 05/11/89 for Limerick Generating Station, Unit 2 III) ACRS CONSULTANT REPORTS 1.
CT-1941 Reed memo to Michelson re Comments on the 09/12/88 Maintenance Proposed Rule, the Industry Standard, and on the Subcommittee Meeting of 7/7/88 2.
CT-1942 Corradini ltr to Kerr re Severe Accident 10/25/88 Research Program Partners Meeting on Containment Loading Issues and NRC Severe Accident Computer Codes 3.
CT-1943 Catton memo to Houston re Severe Accident 11/16/88 I'esearch Partners Meeting 10/20-21/88 --
Direct Containment Heating (DCH) i l
C0NTENTS (cont'd)
DATE III) ACRS CONSULTANTS REPORTS (cont'd) 4.
CT-1944 Carter ltr to Igne re Occupational and 04/25/89 Environmental Protection Systems Subcom-mittee Meeting on 4/20/89 re Proposed Interim Standard for Hot Particles l
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May 26, 1989 The Honorable Lando W. Zech, Jr.
Chairman
'U.S. Nuclear Regulatory Commission Washington, D.C.
20555 i
Dear Chairman Zech:
SUBJECT:
THREE HUNDRED FORTY-NINTH MEETING 0F THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS MAY 3-6, 1989 ACRS REPORTS l
The Committee prepared its report on the second draft of NUREG-1150,
" Severe Accident Risks:
An Assessment for Five U.S.
Nuclear Power Plants" (Report to Chairman Zech dated May 9,1989).
The Committee prepared its report on the Operating License Application l.
for Limerick Generating Station, Unit 2 (Report to Chairman Zech dated l.
May ll, 1989).
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i The Committee prepared its report on the Draft Generic Letter Related to Occupational Radiation Exposure of Skin from Hot Particles (Report to Chairman Zech dated May 9, 1989).
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1 The Committee prepared its report on the proposed Generic ' Letter on Safety-Related Motor-0perated Valve Testing and Surveillance (Report to Chairman Zech dated May 9, 1989).
j The Committee prepared its report on NRC's Human Factors Programs and Initiatives (Report to Chairman Zech dated May 9, 1989).
Copies of these reports have been sent to you.
OTHER ACTIONS, AGREEMENTS, ASSIGNMENTS, AND REQUESTS The Committee met with the Commissioners on May 3,1989 to discuss the following issues:
a.
Implementation of the Commission's safety goal policy statement, b.
Proposed NRC maintenance rule, c.
NUREG-1150, " Severe Accident Risks: An Assessment for Five U.S.
Nuclear Power Plants," and
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d.
Integrated approach to regulatory matters.
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- The Honorable Lando W. Zech, Jr.
2 May 26, 1989
. The Comission requested the Comittee's thoughts on how best to inte-grate the regulatory process.
The Comission also requested that the Committee provide its. views on the staff's proposal for the use of l
NUREG-1150 during the period while the report is undergoing peer review.
The Committee prepared a report on the second draft of NUREG-1150 during the May 3-6, 1989 ACRS meeting (see report to Chairman Zech referenced above) based on a limited review of this draft and plans to continue its review of this draft.
The Committee discussed a proposed report to the Comission regarding the status of the NRC staff's implementation of the ATWS rule.
The Committee did not complete its deliberations and plans further consideration of this report during its June 8-10, 1989 meeting.
The Committee agreed to attend the second international meeting of nuclear reactor safety comitte?s which has been proposed by the GPR and RSK..The meeting is to be held in Strasbourg, France sometime during the period of April-June 1990.
The last meeting between these groups was held in Racine, Wisconsin during October 1986.
Since the last report of ACRS activities, the following subcommittee meetings have been held:
Improved Light Water Reactors, April 11-12, 1989
'Jr.
Subcommittee continued its review of the EPRI ALWR Requirements Document.
Containment Systems and Structural Engineering, April 18, 1989 -
The combined Subcommittees discussed current containment design and performance criteria and proposals for containment design and performance criteria for future plants.
Human Factors, April 19, 1989 - The Subcommittee reviewed the Human Factors Research Program Plan and NRC human factors initiatives.
Occupational and Environmental Protection Systems, April 20, 1989 -
The Subcommittee discussed proposed interim standards for occupa-tional exposure of the skin to beta radiation from small particles (hot particles).
Instrument and Control Systems April 21, 1989 - The Subcommittee discussed the status of the implementation of the ATWS rule.
Limerick Unit 2, April 25, 1989 - The Subcommittee reviewed the application of Philadelphia Electric Company for a license to operate Limerick Unit 2.
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t t The Honorable Lando W. Zech,'Jr.
3 May 26,.1989-Materials and Metallurgy, April 27, 1989 - The. Subcommittee' dis-cussed. the status of the ongoing work - on erosion / corrosion of pipes, hydrogen / water chemistry, and zinc addition to primary coolant loor, ater and its effect on materials.
Mechanical Components, May 3, 1989 - The Subcommittee' continued its review of the NRC staff's proposed requirements on the-testing and surveillance of all safety-related motor-operated valves.
FUTURE AGENDA The Committee agreed to.the following schedule for the June 8-10, 1989 ACRS meeting:
Implementation of ATWS Rule - Discuss a proposed ACRS report to the NRC regarding the status of the implementation of the NRC rule on Anticipat-ed Transients Without Scram.
Education Requirements for Senior Operators and Supervisors at Nuclear Power Plants - Review and report on proposed NRC. rules (10 CFR 50 and
- 55) on Education Requirements for Senior Operators and Supervisors at Neclear Power Plants.
Thermal Hydraulic Research Program Plan - Review and report on the status and plans of.the NRC research program related to thermal hydrau-lic research as detailed in NUREG-1252 and a proposed SECY paper to the Commission.
USI A-47,. Safety Im111 cations of Control Systems - Review and report on proposed final resolution of this unresolved safety issue.
BWR Thermal Hydraulic Instability - Review and report regarding the status of work related to BWR thermal hydraulic instability as evidenced by the core power oscillation event which occurred at the LaSalle nuclear power plant.
USI A-17, Systems Interactions - Review and report on the proposed final resolution of this unresolved safety issue.
Performance Indicator Program - Briefing by NRC staff regarding the development and implementation of new performance indicators for operat-ing nuclear power plants.
Service Water Systems - Review and report on the proposed NRC generic letter regarding the impact of service water systems failures and degradations on safety-related equipment.
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The -Honorable Lando W. Zech, Jr.
'4-May 26, 1989 GE ' Advanced. Boiling Water Reactor - Comments by 'ACRS Subcommittee re-garding design features of this standardized reactor to address severe accident policy considerations.
Sincerely, Forrest J..Remick Chairman 1
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%, 4. +,0*g May 9, 1989 The Honorable L.ando W. Zech, Jr.
Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Chairman Zech:
SUBJECT:
NRC'S HUMAN FACTORS PROGRAMS AND INITIATIVES During the 349th meeting of the Advisory Committee on Reactor Safeguards, May 3-6, 1989, we discussed the draft Commission paper related to the NRC's human factors programs and initiatives.
Our Subcommittee on Human Factors dis-cussed this matter with the staff during a meeting held on April 19, 1989.
The subcommittee previously discussed ' draft Revision 1 of the Human Factors Regulatory Research Program Plan with the staff on January 26, 1989. We also had the benefit of the document referenced.
We are pleased that the NRC again is devoting a portion of its research program to human factors issues.
The list of topic areas being worked on or planned is extensive.
This will require dedicated research program manage-ment attention to help ensure that the research progresses in a timely fashion and the results are provided in a form for possible use by the agency.
During the January 26, 1989 meeting of our Human Factors Subcommittee, it concluded that the Human Factors Regulatory Research Program Plan be expanded into a human factors plan for the entire agency, i.e., to include the human factors programs and initiatives of all of the NRC offices.
We are pleased to see that the staff has subsequently reached the same conclusion.
We believe that the more comprehensive document will be of greater use to the Commission and to the interested individuals.
We recommend that the dis-cussion of the other office programs and initiatives be retained in the NUREG document when issued.
We believe that the Office of Nuclear Materials Safety and Safeguards' human factors initiative, addressing material and fuel cycle activities, is a welcome and needed addition to the NRC human factors efforts.
Because few human factors considerations have been included in these activities in the past, much effort will be required.
It is likely that additional human factors personnel will be needed by NMSS to carry out these activities in an effective manner.
i The utilization of a number of diverse institutions and organizations as human factors research providers is commendable.
This is particularly I
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.e The Honorable Lando W. Zech, Jr. May 9, 1989 noteworthy in the organization and management and in the reliability assess-ment program elements of the research plan.
The use of diverse research providers has already generated new input to, as well as interest in, the human factors research program.
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Finally, we have recommended to the staff that a human factors research
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effort be initiated to develop improved methodology for the selection and training of resident inspectors.
These individuals play a significant role in the regulatory program for operating nuclear power plants.
Effective resident inspectors can have an extremely positive impact on nuclear safety through their interfacing role between the NRC and licensees.
Conversely, inspectors who are poorly qualified either technically or in their approach j
to regulation or their interpersonal skills can have a detrimental impact on nuclear plant safety performance. We believe that appropriate human factors research could develop aptitude testing to assist in the selection of resi-dent inspectors and develop training material relating to their work assign-ments and their relationship to licensee personnel.
We recommend proceeding with the proposed human factors research program and initiatives. We would like to be briefed by the staff en the results of the I
research and any proposed implementation into the regulatory process at appropriate times.
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Sincere Forrest J. Remick Chairman
Reference:
Letter dated March 31, 1989 from F.
D.
Coffman, Jr., Office of Huclear Regulatory Research to Herman Alderman, ACRS, transmitting the Commission
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Information Paper on NRC's Human Factors Programs and Initiatives i
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May 9, 1989 The Honorable Lando W. Zech, Jr.
Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Chairman Zech:
SUBJECT:
GENERIC LETTER ON SAFETY-RELATED MOTOR-0PERATED VALVE TEST"G AND SURVEILLANCE During the 349th meeting of the Advisory Committee on Reactor Safeguards, May l
3-6, 1989, we discussed the subject generic letter. This matter was also l
considered during our meetings on September 8-10, 1988, October 6-8, 1988, and April 6-8, 1989.
Our Subcommittee on Mechanical Components discussed this issue during several recent meetings, including one on May 3,1989.
During these meetings, we had the benefit of discussions with representatives l
of the NRC staff and its consultants and the Nuclear Management and Resources Council. We also had the benefit of the document referenced.
We have been following the NRC staff's activities concerning the industrywide roblem of deficient performance of safety-related motor-operated valves p(MOVs) for several years and have held numerous meetings to discuss this issue. We consider the apparent unreliability of such valves and the poten-tial inability of some valves to function under design-basis conditions to be I
I a significant safety issue of high priority.
Further, we are concerned that i
the stroke-timing test prescribed by the regulations is not a valid test of valve operability under design-basis conditions.
This is a test that con-I sists of stroking the valve open and closed, usually without flow or elevated pressure, and measuring the stroke time.
Because of similar concerns, the staff issued Bulletin 85-03, which required a special operability assurance program for certain MOVs in two high-pressure safety systems (i.e., high-pressure injection and auxiliary feedwater sys-tems).
This program was to ensure that the switch settings for the motor operators on these valves would be selected, set, and maintained so that the valves will be capable of performing their intended design-basis functions for the life of the plant.
The staff is now preparing to issue a generic letter to extend the scope of Bulletin 85-03 to all safety-related and position-changeable MOVs.
It also suggests that other MOVs in the balance of plant be considered for inclusion, commensurate with the licensee's assessment of their importance to safety.
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The Honorable Lando W. Zech, Jr. May 9, 1989 9
We' concur with the need for and scope of the proposed generic letter in order to formalize a program to deal with this problem.- However, we.believe that
.it should be revised to incorporate the following recommendations:
1.
The matter of MOV testing and surveillance should be approached in two stages.
For the first stage, each licensee should perform a review and develop documentation of the existing approved design basis governing the selection of each MOV and establish the " correct" MOV switch set-tings using the best available data and calculational' methods.
The in-plant MOV settings should be changed to the selected values and the design-basis operability demonstrated, to the extent practical, using in situ tests and state-of-the-art testing procedures, extrapolation techniques, and diagnostic equipment.
For the second stage, each licensee should complete its program for demonstrating the operability of each M0V by testing under design-basis conditions (preferably reflecting current regulatory requirements as noted in item 2 below) or by using an acceptable alternative.
This may require' extensive out-of-plant prototype testing and analysis.
2.
Although no change in the existing plant design basis is intended by the generic ' letter, we recomend that each licensee be reminded to review the design basis governing the selection of each MOV from the viewpoint of completeness and adequacy in light of current regulatory require-ments.
In the meantime, and to the extent possible, current require-ments should be reflected in selecting MOV switch settings and demon-strating operability.
3.
The present draft of the gene'.
letter appears to permit alternatives to in situ design-basis testing only if it is precluded by the existing l
plant configu ation. We consider this requirement to be too restrictive l
and recomend that reasonable alternatives t;e permitted at the option of the licensee even if in situ testing is possible.
4.
The generic letter does not clearly state the circumstances under which a demonstration of operability (e.nder design-basis conditions might need u
to be repeated in the future g., after a major maintenance or modi-ficationisperformed). We believe that this needs to be clarified.
Our intention in recommending a two-stage approach is to encourage an early implementation of the imediately achievable portions of the generic letter while work proceeds on a reasonable schedule to develop the required calcu-lational and testing capabilities and to comp 1cte the tests required to achieve full implementation.
We believe that a two-stage approach will ensure a more orderly achievement of the objectives of the generic letter.
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h Ihe Honorab'le Lando W. Zech, Jr.
- Priority consideration'should be given in both stages to those MOVs that the licensee considers to have the greatest. impact on plant safety.
Sincerel orrest J. Remick Chairman
Reference:
Letter cated April 26, 1989 from E. S. Beckjord, Office of Nuclear Regulatory Research, to T. E. Murley, Office of Nuclear Reactor Regulation,
Subject:
Transmittal of. Generic Letter on Motor-Operated Valve Testing and Sur-veillance I
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5 May 9, 1989 The Honorable Lando W. Zech,-Jr.
Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Chairman Zech:
SUBJECT:
GENERIC LETTER RELATED TO OCCUPATIONAL RADIATION EXPOSURE OF SKIN FROM HOT PARTICLES During the 349th meeting of the Advisory Committee on Reactor Safeguards, May 3-6, 1989, we reviewed the referenced draft generic letter, including a draft Interim Standard on Occupational Dose for Skin from Beta Radiation Emitted from a Hot Particle.
Our Subcommittee on.0 occupational and Environmental Protection Systems, its consultants, and invited expert,'Dr. Dade W. Moeller, discussed this r;atter during a meeting held on April 20, 1989 with represen-tatives of the NRC staff, the National Ccuncil on Radiation Protection and Measurements (NCRP), and the Nuclear Management and Resources Council (NUMARC). We also had the benefit of the documents referenced.
During the past few years, high sensitivity personnel contamination monitor-ing equipment has been installed in most nuclear power plants to improve their radiation protection programs.
This has resulted in the occasional discovery of microscopic hot particles on workers' skin and clothing at many nucitar power. plants.
(Fragments from Stellite faced components containing cobalt-60 and irradiated fuel fragments are the most common hot particles.)
It is clear.that het particles have always been around nuclear power plants f
but generally were not detected. We have been told that there is no evidence i
that these hot particles have caused workers any adverse health effects. The staff has concluded that the existing 10 CFR Part 20 limits intended for exposures of large areas of skin (7.5 rem per quarter for skin of the whole body and 18.75 rem per quarter for the extremities) are overly restrictive when highly-localized exposure results from a hot particle.
The staff plans to amend 10 CFR Part 20 to provide a less restrictive limit for exposure of the skin by hot particles.
Until this amendment to 10 CFR Part 20 becomes I
effective, the staff proposes to use the interim standard, that is enclosed in draft form with the generic letter, in taking enforcement actions.
L Industry representatives have been expressing concern since 1987 that, as a result of the current interpretation of the regulation, an unduly high level of attention and emphasis is being given to hot particle doses at nuclear power plants.
These representatives have indicated that this situation is causing unnecessary fear and concern among nuclear power plant workers.
We believe this to be a very serious issue.
Industry has also provided data showing that workers could be exposed to substantially less whole-body a
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l The Honorable Lando W. Zech, Jr. May 9, 1989 radiation (from sources other than bot particles) by setting a more realistic hot particle exposure limit.
In order to avoid what the staff is considering as " overexposure" from hot particles, licensee radiation protection programs require that workers be monitored frequently for hot particles during work in areas that have the potential for hot particle exposures. This more frequent,
monitoring increases the time workers spend in radiation areas tc complete a given task and thus increases whole-body radiation exposures. The results of an industry survey reported by NUMARC indicate that implementation of a more realistic limit (discussed below) for hot particle exposure would result in an estimated reduction in whole-body dose of 5 to 45 person-rem per year per nuclear power plant unit.
(For 1987, the average total collective dose per unit was 420 person-rem.)
Other concerns expressed by industry are cost related (reduced worker pro-ductivity and the need for more health physics technicians), increased radwaste volume, impact on SALP ratings, and potential insurance and legal considerations.
Industry representatives have emphasized that a change in the NRC position would not result in a decrease in the protection of workers or the general public nor in the controls that have been established to prevent hot parti-cles from being transported off-site.
The staff, in March 1987, asked the National Council on Radiation Protection j
and Measurements (NCRP) to study the health significance of exposure from hot particles on the skin and to provide recommendations based on the findings of this study.
(NCRP has an international reputation for excellence in the field of radiation protection and has been chartered by Congress to work with matters.)gencies and others in developing guidance in radiation protection federal a A five-person NCRP subcommittee made this study, and the NCRP provided a report entitled, " Recommendations on Limits of Exposure to ' Hot l
Particles' on the Skin" to the staff on June 17, 1988.
This report was l
subsequently reviewed and approved by the full 75-member NCRP.
The NCRP recommendations are " based on ensuring that ulceration of minute areas of the skin" does not occur. The risk of radiation-induced skin cancer from exposure to a hot particle was not considered to be significant or controlling by NCRP.
NCRP's recomended exposure limit for particles less than 1 m in diameter is IE+10 beta particles emitted from the surface of the particle.
(This limit is expressed as 75 microcurie-hours where one beta particle is emitted per disintegration.) They recomend that any overexposed individual be provided with follow-up medica? evaluation with respect to skin ulceration.
Depending on particle size and isotopic composition, this results in a dose limit ranging from 300 to 800 rad.
To place this dose in perspective, a 2000 rad dose is the accepted limit for radiotherapy treatment involving large areas of the skin. This limit is also based on avoiding skin ulceration.
In its June 17, 1988 transmittal letter, NCRP stated that its recommendations l
may be considered " firm" (subject to final editorial changes) and "may be
9 The Honorable Lando W. Zech, Jr.
- May 9, 1989
__used and quoted as appropriate." This letter indicated that the NCRP report would be published-in final form in the fall of 1988. The staff subsequently raised a number of technical and philosophical questions with respect to the NCRP recommendations that are currently in the process of being answered.
NCRP also requesten aat NUMARC provide coments on. the NCRP report.
NUMARC's comments _ supported NCRP's approach to the hot particle problem but pointed out what NUMARC believed to be considerable conservatism used in'the NCRP recommendations.
As a result of the staff and NUMARC coments, there is no firm schedule for final publication of the NCRP report.
The staff plans to revise appropriate sections of 10 CFR Part 20 to limit hot particle exposure of the skin and will consider the final NCRP recomenda-tions-and recent research results.
However, the staff recognizes that it will be at least two years until this revision esn become effective and believes that it is appropriate to use an interim standard in the exercise of its enforcement discretion regarding hot particle exposures.
The staff considered implementing the recommendations in the NCRP report in its interim standard for skin exposures to hot particles. However, the staff decided, for a variety of reasons cited in the draft generic letter, that it would be inappropriate to implement these NCRP recommendations at this time.
Instead, the interim standard enclosed with the draft generic 'etter, in effect, changes the limit for exposure of the skin to radiation from hot particles from 7.5 rem (skin of the whole body) or 18.75 rem (skin of the hands and forearms, and feet and enkles? per calendar quarter to 50 rad per hot particle exposure.
Recommendations We do not endorse the staff's proposal to issue the generic letter and interim standard in its present form.
Industry, in its presentation to us, has made a strong case that the proposed interim standard for hot particle exposure would provide very little relief in addressing the hot particle problem and believes that the interim standard should be based on the NCRP recommendations.
The staff, on the other hand, has obvious difficulty in basing an interim standard on an unpublished NCRP report. Accordingly, we recomend that staff senior management take an active role in effecting a timely resolution of remaining outstanding issues with NCRP so that its report may be published.
The staff should then develop on an expedited basis an interim standard based on the NCRP recommendations.
Based on what we have been told, we believe that this interim standard could be completed by September 1989.
To the extent the standard differs from the NRCP recommendations, the staff's reasons for such modifications should be clearly and completely documented.
Also, the staff concurrently should move ahead with its planned revision of 10 CFR Part 20 rulemaking on this subject.
The Honorable Lando W Zech, Jr. May 9, 1989 There are two additional items concerning the draft generic letter and interim standard that we believe should be corrected in the final interim standard.
First, the draft interim standard fails to define a hot particle with respect to size for purposes of regulatory control.
This is a very important issue, since the size of the exposed area of skin is central to the determination as to whether the exposure limits for large areas of skin or hot particles should be used.
NCRP uses 1 millimeter as the maximum size that should be used in implementing its recommendations.
We believe that this issue needs to be clarified in the final version of the interim standard and in the planned revision of 10 CFR Part 20 on hot particles.
Second, we recommend that the regulatory concept contained in Section 4, Occupational Exposure Limit, of the draft interim standard be reconsidered.
The section states that the NRC will not issue a notice of violation (NOV) for a single hot particle exposure (less than the proposed limit) to an individual during a calendar quarter.
It further states that the staff may issue an NOV if any individual is exposed to two or more hot particles during a single event or to hot particles in two or more separate events during a calendar quarter. This policy appears to be an unnecessary and complicating feature of the draft interim standard given the existing regulatory require-ments of 10 CFR Part 20.201, Surveys, which requires that licensees must perform " adequate surveys."
It is also inconsistent with the staff's posi-tion that hot particle exposures are not to be added to skin dose for record-keeping purposes and are not themselves additive unless they occur in the same location on the skin.
We intend to follow the progress of the interim and final resolutions of this difficult and controversial issue and will provide you with further coments I
as appropriate.
Sincerel,
orrest J. Remick Chairman
References:
1.
Letter dated February 9,1989 from J. H. Sniezek, Office of Nuclear Reactor Regulation, to E. L. Jordan, Committee to Review Generic Re-quirements,
Subject:
Generic Letter and Interim Standard Concerning Hot l
Particle Exposures of Skin l
2.
Letter dated June 17, 1988 from W. R. Ney, National Council on Radiation Protection and Measurements, to R.
E.
Alexander, Office of Nuclear Regulatory Research, transmitting NCRP Report 80-1, "Reconnendations on Limits of Exposure to ' Hot Particles' on the Skin" (draft of June 1988/Rev. 3)
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The Honorable Lando W. Zech, Jr.
Chairman
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Dear Chairman Zech:
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SUBJECT:
NUREG-1150, " SEVERE ACCIDENT RISKS: AN ASSESSMENT-FOR FIVE U.S.
NUCLEAR POWER PLANTS" During the=349th meeting'of the Advisory Committee on Reactor Safeguards, May
'3-6, -1989, we discussed the second draft of NUREG-1150, " Severe Accident b
. Risks:
An Assessment for Five Ll.S. Nuclear Power Plants," with members of the staff. We also had.the benefit of the documents referenced.
Although we have not had an opportunity for more than a brief look at this 1
second draft. we have been asked to recomend uses to which it could he put
, before the completion of'the peer review as organized by the NRC staff.
At this time, on the basis _ of a cursory examination, we can recomend only that, if its conclusions are used, they should be examined very carefully in light of the criticisms leveled at the initial draft. 'For the most part, criticism of the. initial draft focused on what has come to be called the Level II portion of the probabilistic risk assessments (PRAs) discussed in the report.
It would appear on this basis that prior to peer review of this second draft, information and insights that may come from the Level I portion of the report can ' be given more credence than those from 'other parts of the PRAs.
We.
observe, however, that the core-damage frequencies reported do not take into account a number of external accident initiators that in other contemporary K
- PRAs have appeared as major contributors to the risk calculated.
Of some interest to us, in connection with staff usage, are coments from some' segments of the staff that might be expected to use either the results or the insights derived from the report.
During the past month we have observed the following:
During our April 6-8, 1989 meeting, the Director of the Office of Nuclear Reactor Regulation reported on a major effort being con-sidered to reduce the risk that he believes is associated with the interfacing-systems LOCA.
We observed that the draft NUREG-1150 report did not identify this as a major risk contributor.
He responded that he was skeptical of the results of PRAs.
He felt I
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The Honorable Lando W. Zech, Jr. May.9, 1989 s',
that,-if his current concerns are borne-out by further investiga-tion, this, issue.is important enough that it 'should be resolved before the individual plant examination (IPE) program is completed.
Also during our April 6-8, 1989 meeting, we discussed with members of the staff from, the Office of Nuclear Reg)ulatory Research the performance of motor-operated valves (MOVs in nuclear power plants.
They presented a study, performed at Brookhaven National Laboratory, which they are using as partial justification for
' requiring a major program of testing, maintenance, and repair of MOVs in operating plants.
The report concludes that the core-damage frequency for boiling water reactors (BWRs), taking into account what they now believe to be the performance of MOVs, is more than an order. of magnitude greater than the core-damage frequency for BWRs reported in the draft NUREG-1150.
On the basis of the staff's conclusion regarding this matte.r, they are recom-mending an extensive program which they believe will enhance valve performance.
They consider this problem so important that it too should not wait for the IPE program.. They are convinced that NUREG-1150 does not represent properly what they view as a major risk contributor.
We conclude from these experiences that it may be worthwhile, in the review process, for those responsible for NUREG-1150 to solicit coments from other elements of the staff which might be expected to use the results of the report.
In sumary, on the basis of a very preliminary review, the insights and the results of the second draft of NUREG-1150 should be used with considerable caution before the-planned peer review has been concluded.
We expect that more credence might be given to the Level I parts of the PRAs than to Levels II and III. However, we repeat that some of the Level I results have already been called into question by other parts of the staff.
Sincerely Forrest J. Remick Chairman
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The Honorable Lando W Zech, Jr. L May 9, 1989
References:
1.
U.S. Nuclear Regulatory Comission, NUREG-1150, " Reactor Risk Reference Document," Volumes 1, 2 and 3, Draft issued for coment, dated February 1987 2..
U.S. Nuclear Regulatory Comission, NUREG-1150, " Severe Accident Risks:
An Assessment for Five U.S.
Nuclear Power Plants," Volumes 1 and 2 (Second Draft for Peer Review), dated April 17, 1989 (Pre Decisional) 3.
Memorandum dated April 18, 1989, for the Commissioners from V. Stello, Jr., Executive Director for Operations, SECY-89-121.
Subject:
Transmit-tal of NUREG-1150, Second Draft for Peer Review 4.
Memorandum dated February 17, 1989, for the Commissioners from V.
Stello, Jr., Executive Director for Operations, SECY-89-058,
Subject:
Status Report and Preliminary Results of NUREG-1150 5.
Memorandum dated December 8,1988, for the Commissioners from V. Stello, Jr., Executive Director for Operations, SECY-88-337,
Subject:
Plans for Future Review of NUREG-1150
_ _ _ - _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ - _ _ - _ ~
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UNITE 3 STATES g
'8 NUCLEAR REGULATORY COMMISSION o
y
,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS '
l vAsmworow, p. c.rossa May 11, 1989 l
1 The Honorable Lando W. Zech, Jr.
Chairman U.S. Nuclear Regulatory Comission Washington, D.C.
20555
Dear Chairman Zech:
SUBJECT:
OPERATING LICENSE APPLICATION FOR LIMERICK GENERATING STATION, UNIT 2 During the 349th meeting of the Advisory Comittee on Reactor Safeguards, May 3-6, 1989, we reviewed the application of the Philadelphia Electric Company, the Applicant, for a license to operate the Limerick Generating Station, Unit 2.
Our Subcommittee on Limerick 2 toured the facility on the morning of April 25, 1989 and met in the afternoon, in Philadelphia, to consider this application.
During our review, we had the benefit of discussions with representatives of the Applicant and the NRC staff.
We also had the benefit of the documents referenced.
In the ACRS report, dated November 6,1984, to then Chairman Nunzio J.
Palladino, the Comittee comented on the application for an operating license for Limerick Unit 1.
In that report, the Comittee noted that, although the Applicant had requested an operating license for both Units 1 and 2, the Comittee felt that it was not appropriate to coment on Unit 2 at that time because of the uncertain schedule for construction and operation of Unit 2.
Although Unit 2 is being considered for an operating license some four-and-a-half years after the approval of an operating license for Unit 1, the two units have the same rated power level, use the same model nuclear steam supply system, and are generally very similar.
In the course of our review, we discussed management and staffing of Unit 2.
Recent changes in the Applicant's management have resulted in the location on-site of a vice president responsible for the Limerick Station.
A number of those individuals responsible for testing and startup of Unit 2 have gained experience on Unit 1.
This experienced group appears to be conducting a well organized and effective test program and to be accomplishing a smooth transition in the turnover of responsibilities to the crew that will be responsible for operation.
The most recent Systematic Assessment of Licensee Performance (SALP) rating by the NRC staff gives the management of this group an unusually high rating for its
' he Honorable Lando W. Zech, Jr. May 11, 1989 T
performance.
We found no reason to question the experience, training, or capability of the personnel who will be responsible for operating Unit 2.
In the ACRS report, dated November 6,1984, the Committee mentioned that a probabilistic risk assessment (PRA) had been performed for Unit 1.
The Applicant now has its own staff of PRA practitioners who, with some outside assistance, have not only revised the PRA for Unit 1 to reflect changes in the plant and in operating practices, but have also performed a PRA for Unit 2.
Among changes that have taken place since the earlier PRA was per-formed are the installation of vents for the containments of both units and the adoption of Revision 3 of the Boiling Water Reactor Emergency Procedure Guidelines.
The PRA group, in a 1988 update, reports a calculated core damage frequency of 6.69E-6 per year for each unit. This is slightly less than half that calculated when the original PRA was performed.
It should be noted that this does not include any contribution from seismic events which have been a significant contributor in other contemporary PRAs.
It appears that the Applicant's organization is using the insights from PRA in training and in planning their maintenance program.
They intend to use these insights also in the formulation of their accident management pro-gram.
In the course of preparing the organization and plant for operation, the Applicant performed a Readiness Program Assessment "to assess the adequacy of existing licensee programs and processes," and retained a contractor to perform a Readiness Verification Program to provide "a
comprehensive integrated process to assess the design, construction, and operational aspects of Unit 2."
The NRC staff then reviewed both of these assessments.
Both the Applicant and the NRC rtaff reported that these reviews provided convincing evidence that the plant is ready for startup.
In the ACRS report, dated November 6, 1984, the Committee recommended also that Unit I receive special attention in the NRC staff's resolution of the unresolved safety issue (USI) on systems interactions. We recommended also that special attention be given to the identification of any risk outliers associated with seismic events.
These issues are being dealt with generically.
Limerick Unit I has had some difficulty with corrosion of fuel cladding; however, this does not appear to be a serious safety problem.
The Appli-cant proposes some changes in plant equipment and operating procedures which should make the corrosion less likely. Although some insights appear to have been developed that may make the problem less severe or perhaps even eliminate it, the results of applying these insights are not yet available.
For the past several years, it has been standard NRC practice to require extended periods of plant operation at very low power before approving operation at full power. Presumably, this has been done in the belief that it is safer than going more directly to higher power operation.
It appears that if Linerick 2 is approved for full power cperation, the Commission l
d
~
The Honorable Lando W. Zech, Jr. May 11, 1989 will recuire several months of operation at less than full power and that probably two months of this will be at about five percent of full power.
However, we have yet to find anyone on the staff who has done or who knows of any systematic attempt to investigate whether there are any negative effects associated with this practice.
Certainly, the units are not designed for extended operation at, for example, five percent of full power.
And at least one licensee representative recently referred to operation at five percent of full power as being " uneasy," although he did not believe there was anything unsafe about it.
We have no evidence that it is unsafe, but do know of instances in which operation at low flow has produced excessive wear in check valves and in which operation at low power has produced excessive vibration in a feedwater pump not designed for extended operation at low flow.
Our principal concern stems from the memory 'that the operators of the Chernobyl plant, Unit 4, were unaware of the dangers of operation at low power, whereas a careful analysis would have convinced them that this was undesirable.
It appears to us that if the practice of extended operation at low power is to be continued, some systematic search for possible harmful effects should be performed.
We' believe that, subject to satisfactory completion of construction and preoperational testing, there is reasonable assurance that the Limerick Generating Station, Unit 2, can be operated at power levels up to 3293 MWt without undue risk to the health and safety of the public.
Mr. James-C. Carroll did not participate in the Comittee's review of this matter.
Sincerely, Forrest J. Remick Chairman
References:
1.
U.S. Nuclear Regulatory Comission, Region I, Systematic Assessment of Licensee Performance Board Report, Philadelphia Electric Company, Limerick Generating Station, Unit 2, Inspection Report 50-353/87-99, February 22, 1989 2.
U.S.
Nuclear Regulatory Comission, NUREG-0991 Supplement No.
7,
" Safety Evaluation Report Related to the Operation of Limerick Gener-ating Station, Units 1 and 2," April 1989 1
[
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.l The Honorable Lando W. Zech, Jr.
4-'
May 11, 1989 l
I
.3.
. Letter dated January 23, 1989.from Gus C. Lainas, U.S. Nuclear Regula-tory Commission, to G. A. Hunger, Jr. Philadelphia Electric Company,
Subject:
Inspection of Independent Construction Assessment, Limerick 1
Generating Station, Unit 2; Inspection Report Number 50-353/88202 4.-
.Public Statements provided during the April 25, 1989 meeting of the J
ACRS' Limerick 2 Subcommittee from the following:
K a.
Marvin I. Lewis.. Limerick Ecology Action b.
Richard Myers, Citizens' League for Energy Awareness and Resources c.
Ruth Miner, Citizens for Environmental Rights L
d.
Emanuel Mendelson, Citizens for Environmental Rights l
i e.
Phyllis Gilbert, Sierra Club, Philadelphia, Pennsylvania i
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NUCLEAR REGULATORY COMMISSION 3
- i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS f
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September 12, 1988 l
l MEMORANDUM FOR:
C. Michelson, Chainnan Subcommittee on Maintenance Practices and i
Procedures FR0ti:
G. A.
L CRS Consultant l
SUBJECT:
COMMENTS ON THE MAINTENANCE PROPOSED RULE, THE I DUSTRY STANDARD, AND ON THE SUBCOMMITTEE MEETING 0 6UL 7,
1988 e.
- 2 W t
a Since these corrents will likely follow by several days the full committee l
decision and letter, and since I expressed my points orally at the meeting, I will be brief for the record as follows:
1 1.
The " expectations" and "shoulds" for an Industry Standard as listed on l
pages 6 and 7 of Enclosure 1, are quite general, and the NRC staff l
people (including Resident Inspectors) can interpret to create moun-l tains of paper to negatively impact the industry costs and safety.
The NUMARC representative indicated they expected to use their exist-ing guideline documents for the most part for the Industry Standard.
There should be a prompt meeting of minds to resolve NRC "expec-tations" vs. NUMARC-INPO. Personally, I feel INPO is burdening the plant people more than enough right now; progress is being made, and the NRC should not chenge or magnify present directions and in-i clusions.
2.
The NRC staff showed slides and furnished documents that clearly addressed nuclear utilities with " poor maintenance programs" as the target of the rule. Upon questioning, the staff stated the word intended was " performance" not programs. This might be followed to see if future paper gorrects wording to " performance" since if the
" programs" wording continues, and is so interpreted, mountains of paper can result even where a utility's perfonnance excels.
3.
At one point in the NRC presentations, a speaker remarked that if l
industry comes up with a good standard then the NRC staff would write a "one page guide." This seems appropriate even if the present INPO i
guidelines documents are used.
i 4.
The loose use of the words " BOP" and the issue by some that BOP should i
come under NRC regulations should be clarified. The failure to include auxiliary feedwater systems as safetyrrelated equipment many
)
years ago should not be used as a " door opener" to clamor for BOP l
t encroachment. Where auxiliary feedwater, instrument air, and certain check valves are necessary to provide for certain decay heat removal service (or other vital safety need) the equipment should be L
\\
classified as safety-related, not BOP.
If a Maintenance Rule, Stan-dard, or whatever, envisions implementation into such as turbine l
maintenance then mamoth amounts of paper will result and all the cost L
benefit results in the enclosures to the V. Stello memorandum to the Commissioners (undated) concerning a proposed maintenance rule will become grossly incorrect.
i 5.
I endorse the Option 3 route and believe the maintenan& issue has been stirred up so much that some sort of rule has to be promulgated for more or less political reasons. Technical justification for a rule doesn'. seem to exist at this point in time.
I would hope the rule would be carefully worded so as not to impact negatively on the efforts under way, or to provide ammunition for paper and procedural mountains vs. real craf tsmanship application, 6.
In my opinion, about one-half of all maintenance activity thus far in the budding nuclear field has been caused or related to flawed design.
(Even the latest AEOD sumary report on incidents stated design was responsible or involved in about 501.) We all know that designers have escaped their appropriate responsibility by the " sole utility licensee" dump off, but we should continue in formulating new rules, guides, etc., to search for avenues to bring designers (vendors, component manufacturers, etc.) to accountability. Apparently in the study and creation process of this proposed rule, the issue' of " des.
ionated representatives" surfaced, but not in very good focus, and not too much cirected toward vendor-manufacturer designers.
Regrettably, utility owners and their maintenance people have struggled long, and will continue to struggle to try to correct flawed design in the absence of forthright involvement of real " designated representatives" from the vendor-manufacturer designers.
The ACRS subcommittee should note that not once does the words " des-ignated representative" appear in the meeting review package or the meeting handouts.
7.
Another word missinc from the paper is the word " selection." Lots is said about program indicators, training, etc., but the basic elementa-ry key to achieving fine maintenance is the selection of personnel who have the appropriate mechanical comprehension and manual dexterity.
This is just absolutely so for I&C maintenance.
I can never forget the early days of a nuclear project where selection was not practiced and where one worker broke enough equipment to keep about two dozen busy repairing same. The ACRS subcommittee might also reflect on TVA performance and the fact that TVA had not used selection techniques, and at least one TVA management person stated "give mrt wann bodies and I can train them to do the best."
8.
The use of " reliability centered maintenance" techniques is somewhat promoted in the staff documents.
In my opinion, it is.nuch too early in time to consider applying reliability centered maintenance to nuclear. There are only about 100 nuclear plants in the U.S.A.,
nothing is standardized, and the variety of components and manufactur-ers are in the thousands. Without standardization, repetitive use of eiactly the same equipment in like situations, a data bank cannot be
e.
I meaningful toward any useful reliability centered maintenance. The fact that the French may be using RCM only supports its nonapplicabil-ity here, since their nuclear is at least somewhat standardized.
9.
It would be unfortunate if maintenance gets so regimented and "verba-tim" that shift operators cannot tighten the packings"Of a manual valve.-- vs. being regimented into a Maintenance Pequest route that takes 10 days before a maintenance worker might approach the valve --
after, of course, the MR has gone through Appreval. Investigation.
Prioritization, Planning, Scheduling, CHAMPS, etc., and finally to physical action.
By this time under regimented and verbatim compli-ance, the valve packing would have been blown and the unit tripped.
10.
Regrettably, " craftsmanship." and motivation are being lost or de-stroyed in the S.S.A.
Since their destruction has proceeded 50 far, particularly in large companies and in certain geographical areas, there is probably no other way than to implement more paper regimen-tation for maintenance at many nuclear utilities. Hopefully, this implementation will not be forced acrnss the board and at utilities where "craf tsmanship." motivation, and performance exist.
If such forcin5 occurs, the good performer will be torn down to the level of the poor, and I doubt that the poor performer will improve drastically with his implemented paper mountains, unless other more astute steps are taken.
11.
In preparing for this meeting, I visited a plant and had telephone communication with a seccnd plant.
The first plant has two units and a total T/0 of about 300 pecole, of which 68 are assigned to (Mech.
Elec. and 180), maintenance supervision, and physical work. This plant has an outstanding performance record (nationally recognized --
not necessarily NRC) with a unit forced outage of about 0-1 per year.
This facility practices "craf tsmar. ship" and is structured to do 50.
The second facility of one unit has about 1400 total people with 157 in maintenance and now practices maintenance paper much in line with what the NRC staff is proposing. The performance record at this second plant is poor by comparison to the first.
It should be noted (and reflected upon) that in lir. Walter Scott's presentation (and handout) in the "Sumary of Recommendation" that the German maintenance qualification and motivation emphasis is on
" craftsmanship." Unfortunately, the name of the nuclear improvement game in the U.S. A. since the THI-2 inciQnt, has been " throw numbers of bodies" at the job.
" Numbers" of people without " quality" of people serves to decouple the striving for excellence in nuclear plant performance.
- 12. At the meeting, an NRC staffer stated that if a good performance was being impacted negatively (referring to costs) because of forced implemented programs from the Rule or Guide, then he could avail himself of protection under the Backfit Rule.
I would like to believe this is so, but doubt such opportunity would exist or be used once the avalanche starts.
_-___________-_____a
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13.
In my opinion, the NRC staff is indulging in too many team ins and has too many planned for maintenance, and in other areas. pections l
When l
one considers all the INPO inspections, there is hardly any time.left at the plants _for the key personnel to focus on what they should be focusing upon.
This is a generic problem of U.S. A. nuclear and.is negatively impacting plant personnel attention, motivation, and performance.
Also it is contributing to an unfortunate key and experienced labor turnover, and the seeking of early retirement.
14 C'est la guerre!!
cc:
R. F. Fraley i
0, Wylie J. Carroll 8
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Department of Nuclear Engineering and Engineering Physics
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University of Wisconsin
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'1530 Jones:9 %e Ma:s:9 WI 527:5-1657 Pnone'E El 253 1646 October 25, 1988 f.E cs.; -
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.. t Professor William Kerr
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Nuclear Engineering Department DI University of Micnigan North Campus
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Ann Arcor, MI 48109
Dear Siil:
I attended tne Severe Accident Research Program Partners meeting last.
week on Thurscay and Fricay.
During those two days containment' loading issues were discussed along with presentations on the NRC severe accident computer coces applied to' speed.fic issues (e.g., SCDAP. MELPROG, CONTAIN, MELCOR). My general impression is that much of the severe accident researen efforts have now been focused on current licensing questions.
This is a deviation from past practices, and seems to be a good response to past comments. This emphasis can have a snort term benefit (1-3 years) for specific IPE issues.
However, because of the shrinking research budget much of the long term issues have to be neglected.
This is an unfortunate situation and presents a " catch-22" that has to be considered.
I would suggest the sub-committee discuss this situation because it appears it may get worse before it gets better.
Let me comment on each of the subjects separately, because there are specific technical points to be made.
I will try to link these various subjects at the end of the discussion.
Hydrocen Combustion The Sandia researchers presented recent results in the area of hydrogen combustion.
It is apparent how extensive the research progress has been over the years.
In the first part of the presentation it was emphasized that high-temperature combustion research was needed in the temperature range above 200*
C; specifically determine volumetric oxidation rates and hydrogen-steam-air detonability and flammability limits.
This is motivated by direct heating phenomena.
This recommendation may seen reasonable, but I am a little dubious as to the uniqueness of this needed data.
At these temperatures and densities one is approaching conditions within an internal combustion engine. There-fore, I would first suggest that an extensive literature review verify that relevant data does not already exist, before an experimental program is begun.
Also I would wonder if we are looking at containment atmospheric conditions where failure probabilities are approaching one.
Therefore, the need for more certainty may not be justified.
Professor William Kerr Page Two Octooer 25, 1988 In the second part of the presentation hydrogen mixing experiments at HDR were predicted by three different containment modelling tools; HECTR, CONTAIN and MELCOR. This was interesting for two reasons.
First, it showed how all tne modelling tools successfulij predict the data.
Second, it graphically illustrates the duplication in cratainment modelling that has developec over the years.
All tnese tools are caite similar in their level of detail and s o;ni sti c ati on.
MELCOR can also model the RPV and thus it is still confusing to me why all these codes are still supported by NRC. HECTR and CONTAIN snoulc be retirac, all the resources focused on MELCOR.
It is so obvious that I wonder why It has not been done yet.
Di rect Heatinp Di rect heating research is a major f ocus of the NRC severe accident research program.
Over this last two years the focus of this research has been on quantifying the fuel dispersal threshhold, the possible mitigative effect of internal containment structures and more recently, the low pressure cutoff for fuel melt dispersal.
Even though this focus has become more practical I am still surprised that the possible mitigative effect of the water coolant on the DCH process has not been systemically investigated.
Early in 1982 tne EPRI sponsored research program at Argonne (entitled CWTI tests) demonstrated the possiDly mitigative effects of water for a 1/30 scale cavity-containment of Zion.
This work was in support of the 1981 Zion Safety Study.
Over fif teen tests were performed that clearly showed the mitigative effects of water at small scales with prototypic materials.
The questio6 of scale must be addressed, but I am surprised that there is no new data afte'-
six years, at any scale.
The BNL presentation emphasized that they are ceveloping a general dispersal criteria as a function of towing pressure for dif ferent cavities with simulant materials.
However, to do this in the i
absence of a sinulant coolant (e.g., LN2) seems to be misguided. Water will be present in many (most) cavity designs and may be an accident management strategy in all cases. To develop a low pressure cutoff without considering it may result in misleading results.
Mark-I Li ner Attack Recently, short term research work has begun at SNL and BNL on assessment of MARK-1 liner f ailure.
One could divide this work into two categories.
First, one examines the transient spreading of the " fuel" melt as it flows across the drywell floor, to determine if it would reach the drywell liner.
Second, assuming melt-liner contact one determines the likely heat transfer rates between the molten fuel and the steel wall.
One should consider the effect of the coolant in both cases. BNL has used small scale sinulant tests to address the first category with some interesting results.
The data clearly indicates that when a coolant (water) is present with sufficient subcooling (or the melt has little superheat) the " fuel" simulant (lead) does not spread I
Professor William Kerr Page Three j
Octooer 25, 1983 l
as easily but freezes as a lump. The problem is that the tests use simulants that may not be appropriate to demonstrate the correct qualitative behavior; e.g., there is no gas release f rom the simulant drywell floor.
Such details must be critically examined to verify that the observations may be valid for larger scales witn real materials The second category of melt-liner heat transfer is being considered by SNL as well as BNL.
The key point in this case is tnat immediate crust growth on the steel liner protects it from direct melt contact.
i This is seen in the BNL simulant tests as well as the small I
scale prototypic tests at SNL. The real question is what is the melt pool temperature nistory, tecause if the melt pool temperature remains high the i
I trust will diminisn in size and large convective heat transfer rates from melt to liner will quickly threaten its integrity.
Molcen C:re-Concrete Interactions Tne SNL researcn group cresented the recent results from the SURC-4 test whicn involved the interaction of a steel and zirconium pool (200 and 20 kg respectively) witn a cry basaltic concrete cavity.
The results are not significantly cifferent form BETA or past SNL tests before zirconium was added, but showed interesting behavior when zirconium was present.. A large temperature excursion was noted which can only.be explained by rapid condensed phase cnetical redox react'.ons between zirconium and the other oxygen bearing r.hemical specier bo;idet the decomposition gases.
Current models do not consicer tnis because layer mixing is not considered.
However, this result is consistent with past BETA observations and must be considered for early time fission product release and gas generation.
These same types of experiments are now ceing conducted by ANL sponsored by EPRI where the melt is primarily oxicic.
Tne most recent test (L-5) is still being analyzed.
A major question in this area again revolves around the possible beneficial effects of water addition on the melt quenching and coolability.
These tests are in the SNL matrix but much further into the future. I feel the priority f or such tests is quite nign.
Severe Accident Code Acolications The many NRC computer models that have been developed are now being used to answer specific regulatory questions. This can have a beneficial result f or the short term.
However, there is still a great deal of duplication in some code development efforts.
Therefore, given the cost of comparison to experimental data, and given the cost of code maintenance some cifficult decisions will have to be made as the research budget contracts.
I do not see any direct evidence of this.
- Pfofessor William Kerr Page Fou r Octocer 25, 1988 If you have any questions, please feel free to give me a call.
Sincerely,
b.
Michael L. Corradini Nuclear Engineering and Engineering Physics Mechanical Engineering MLC:1cw
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TO:
Dean Houston 16 Novembsr~1'988
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FROM:
Ivan Catton M
ADW1501 M I
REACTOR SAFEGUN SUBJECT Severe Accident Research Partners Meeting Bethesda, Maryland NOV 211968 20-21 October 1988 ME gg74Gibb%M I attended the final two days of the above referenced meeting.
Material I was interested in was scheduled at e beginning or I chose the part of the meeting where Direct Containment Heating ge
'.,h (DCH) was to be discussed. There have been several code studies of the core degradation process and the impact of natural 4
circulation. The results point to DCH being much lessor' issue than SNL would lead one to believe. The SNL experimental and analytical studies are still based on too much molten core.
The issue of how much, how hot and how fast still does not receive the attention it deserves. The Mark I liner failure issue suffers from the same problem. Convincing arguments were given that the liner will fail when large fractions of the core must be dealt with. More detailed discussion of the meeting topics follows.
Dana Powers discussed a number of issues related to DCH. He made the point that no containment i s i mmune. The source term was not recognized in WASH 1400 and as a result has not been properly quantified. To quantify the source term one must first determine how much energy gets into the containment (quantif y debris dispersal from the cavity and chemical reactions with the atmosphere) and second determine how much hydrogen production will occur. Surtsey (1/10 scale) experiments were conducted to try and understand some of the physical processes important to DCH. The first few experiments have established that increasing mass of the injected molten materials resulted in a decreasing energy release per unit mass. This is not a surprising result although it is certainly comf orting to see it based on experiment rather than supposition. Contrary to the Bob Henry views, it was f ound that structures in the flow path can enhance energy transfer to the containment atmosphere. It is not clear how this conclusion translates to a full scale PWR as most of the structure referred to in the past is in the keyway or corridors leading from the reactor cavity to the containment volume.
Models being developed to predict the DCH induced containment pressure are CONTAIN and KIVA. CONTAIN is a lumped parameter code that has been around for a l ong t i me. It is interesting that the code that was to address issues like DCH was at one time to be HECTOR (sp?). It appears that HECTOR has fallen from favor. KIVA is claimed to be a detailed hydrodynamics code based on finite difference methods that will do \\ mechanistic analysis.
Considering our ability to predict such phenomena, I view the use of KIVA as fun and games. CONTAIN has been augmented with models to describe debris clouds, structural interactions and chemistry. It was claimed tnct the codes do reasonably well in predicting DCH.
2-Accident management studies are underway at SNL. Depressurization i evaluation led to the conclusion that one must get to very low" pressures to achieve beneficial effects. The experiments show that there is a critical pressure where core debris dispersal goes from essentially zero to all of it. Without understanding the physics one cannot begin to guess what the effect of scale will be. An interesting bottom line was reached: DCH impact is small if less than 20% of the core is involved. The basis for this was not clear.
Further there was no discussion of the rate of dispersal.
SNL has still not looked caref ully at the initia* conditions for DCH.
Powers presentation seemed to be based on massive amounts of the core and varying fractions being dispersed. It seems to me that CONTAIN is good enough and its time to put our efforts into determining just what it is that we must deal with. We always get back to i) whether or not the primary system will fail before the core melts through the bottom and if it does, ii) how much of it must we deal with, iii) how het will it be, and iv) how rapidly will it be blown into the containment atmosphere. The first item requires that natural circulation induced heatup of the upper parts of the primary system be properly evaluated. This is being studied and is discussed below.
Ginsberg described the 1/40th scale experimental work underway at BNL.
Ginsberg has injected the simulated core materials into a large number of fluids. These fluids include water and liquid nitrogen. I could not establish what simulant characteristics were used to select the various combinations. It seems to me that energy transfer to the atmosphere is dependent on the drop size of the atomized core materials which means the Weber number is an important parameter yet it was not mentioned.
High pressure blowdown usually means Mach number should be another important parameter yet it was not mentioned. Energy transport and chemistry also depend on the Reynolds number (based on the relative velocity). It also does not seem to be a part of the scaling-analysis.
Either I did not understand what I was being told or the BNL program is just another group of nuclear engineers having fun. Work like that described can be an Important part of developing the needed models for prediction in codes like contain. One must, however, pay attention to scaling and if the conclusions are at odds with the views of others, some effort should be made to enlighten.
Bradley (SNL) described some recent experiments that were to address the Mark I liner failure concern. The experiments were poorly conceived and demonstrated a lack of understanding of the important physical processes.
A series of very nice basic experiments are being conducted by Greene i
at BNL. He has systematically addressed many aspects of the Mark I wall f ailure question. His work should now be used to address the problem. It is interesting that Greene is one of the few involved in, severe accident phenomena studies who consistently publishes his work in respectable archival journal s.
It seems much of the NRC sponsored research is either unpublishable or they do not care. The discussions that followed his presentation put a different complexion on the issue of Mark I liner failure. The liner that may fail is less than ten feet from the pedastl doorway. It was argued that the molten core materials will slosh out and literally run up the wall. The question always gets back to how much, how f ast and how hot. If a large fraction of the core must be dealt with, then further study will not change the conclusion that one must protect the liner.
e v.
Altmeyer from Germany presented the results of a
study of what happens to the molten core on the concrete basemat. He predicts that it will penetrate in about five days or so. Containment evaluation should take this into consideration. To stop the downward pr op a g at i on, one must chill and fracture the debris. Water from above will not do the job as it only fractures a thin surface layer. The j
i results of his study sounded like the China Syndrome revisited. He found that the molten' core penetrated 80 meters into the earth and spread laterally to 30 meters in a period of two to three years (certai nl y lots of time for intervention).
In-vessel cor e melt studies using RELAP5/SCDAP were was described by Allison (INEL). I am continually ama:ed at how well one dimensional codes can describe two dimensional behavior. Allison consicers crust failure (around the molten pool) a major factor in molten pool relocation. How SCDAP addresses this without a detailed pool convection model is truly amazing.
A similar analysis was carried out using MELPROG/ TRAC. It sounded good but Heames (LANL) did not seem to know too much about how the code did it s job.
Natural ci rcul ati on during a TLMB' incident was studied using SCDAP/RELAP5. To accommodate het leg countercurrent single phase stratified flow they modeled the hot leg as two pipes. The core was represented by three parallel channels. The modeling was quite crude but probably adequate to give some indication of whether the lower head would be the vessel failure location. It was concluded that the surge line will fail,.the steam generator tubes will not fail and that the surge line failure will proceed the lower head failure by one hour. Although I would take issue with the conclusion regarJing the steam generator failure, that the failure would not be the lower head is reasonable. A number of sensitivity studies were carried out and it was found that nothing seemed to change the conclusion that the ex-vessel part of the RCS would fail and depressuri e the system before the the lower head fails. Each new series of calculations of in-vessel natural circulation seems to point more convincingly to early failure of the primary system and a decreased probability of DCH induced containment fa:1ure. In this regard, I was pleased to hear that Zuber will be looking into the DCH issue.
l Bergeron (SNL) described his studies of DCH using CONTAIN based on 75% of the core. Although his study was well done, using 75% of the core as an i ni ti al condition prejudices me. He found, in contradiction to Power's interpretation of the Surtsey experiments, that structure mitigates the impact by absorbing energy through heat transfer to it. He f ound that the biggest uncertainty is whether or not hydrogen burns. It was found that debris transport to the upper containment from the reactor cavi ty i s not the dominant mechanism for energy transport. Rather the steam acts as a working fluid. Bergeron concluded by noting that DCH seems to be less I
threatening to large dry PWRs than it was earlier thought to be.
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=O CT-/W MELVIN W. CARTER, Ph.D.
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International RadiatforMrotection Consultant April 25,1989 EELElW.D e.. yv cow.unr On
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Mr. AlIgne APR 2 61989' Senior Staff Engineer
,., ' g Advisory Committee on Reactor Safeguards
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U.S. Nuclear Regulatory Commission Washington, D. C. 20555
Dear A1:
As discussed at the end of our Occupational and Environmental Protection Systems Sub-Committee Meeting on April 20 to consider the proposed Interim Standard for Hot Particles, I'm pleased to send these comments and suggestions. In my opinion, the so-called " hot particle problem" has been blown out of all proportion to any measured or demonstrated worker or public health significance. Thus, as in many other cases, this specific consideration should be balanced with other related considerations and evaluated on an integrated basis.
More than a year ago, I (as a Member of the NCRP) wrote the words quoted below concerning this aspect of hot particles. They were written based on my review of the NCRP Draft Report " Recommendations on Limits of Exposure to Hot Particles". They are taken from the text of my letter to NCRP, dated March 26,1988.
"A major need is to put the hot particle problem into a proper perspective so it can and probably will receive attention and resources proportional to r
its contribution to actual worker health risks. A possible danger in minimizing exposures to hot particles, as useful as it may be, is that such procedures may tend to increase the collective whole body dose received by workers at a given site. These procedures pertain to work practices such as limited stay times, required protective clothing changes at specific intervals, and increased use of manpower to perform hot particle surveys."
I completely support the NRC Staff in its approach to using an Interim Standard for Hot Particles. Hopefully, this process can be compressed in time so the Interim Standard can be issued in order to serve its intended purpose. It is an Interim Standard and should be considered as such in its promulgation and use.
l 4621 Ellisbury Drive, Atlanta, Georgia 30338 (404)458 9474 m
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- Mr. Al Igne c
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April 25,1989 I
1 Brthermore, I fully support the NRC Staff in its comprehensive approach to consideration of the longer-range modification of 10 CFR 20 with regard to hot l
particles. This process should be thorough and deliberate. I also support the concept of using a limit based on the time integral of activity to prevent ulceration of minute j
areas of the skin.
Ma> ccncerns are that enough consideration does not seem to have been given to how the operational heahh physics decision is made as to whether a given exposure is a hot particle exposure or one of an extremity or the whole body. This is important.
1 A similar concern is related to the operational health physics decision as to whether or not one is dealing with hot particles or hot fragments. These are to be excluded i
according to the NRC Staff proposal. However, how is the decision made for I
distinction between hot particles and hot fragments?
If categories of exposure (i.e. hot particle vs. extremity or whole body) and of material (i.e. hot particles vs. hot fragments) are to be determined, operational procedures should be available and documented to support these decisions. It is not clear from the discussions on April 20 that this is the case.
With these reservations, I support the specific Interim Standard for Hot Particles proposed by the NRC Staff. It is a reasonable standard at this time, is relatively l
conservative, and provides some relief for nuclear utility workers while adequately protecting them in their work environment.
I Hopefully, these comments and suggestions will be ofinterest and use to the Sub-Committee on Occupational and Environmental Protection Systems.
Sincerely, 6
Melvin W. Carter MWC/bc
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