ML20049A309
| ML20049A309 | |
| Person / Time | |
|---|---|
| Issue date: | 08/20/1980 |
| From: | Minogue R NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | Schroeder F Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20049A310 | List: |
| References | |
| FOIA-82-118, RTR-NUREG-0660, RTR-NUREG-660, TASK-***, TASK-TM NUDOCS 8011050367 | |
| Download: ML20049A309 (53) | |
Text
b-R G. A. Arletto W. ft. "ocrison E. C. Uenzinger J. L. Milhoan K. R. Goller M. Parsont AUG 2 01980 R. Alexander E. Conti J. Norberg W. Anderson "E"CRANDUM FOR:
F. Schreeder, Acting Director Division of Safety Technology Office of clear Reactor Regulation F30:15 Robert B. !!inogue, Director
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Office.of Standards Develepnent SUEJECT:
SAFETY RATIO! ALE FOR ACTION PLAN
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Sy memorandum of July 11,1980, the E00 requested SD provide to you a safety rationale for all TIII Action Plan items for v.hich we have lead responsibility (as identified in Table 1 of NUREG-0560). Attached are initial drafts for the SD identified tasks. The initial drafts were developed in accordance with the instructions contained in the July 11, 1980 mer.orandum.
Cognizant SD Branch Chiefs for the attached Action Plan ite s are as follows:
\\# E. Uenzinger, RSSB - I.A.I.4, I'.A.2.6, I.A.3.3 I.A.4.2, I.D.4, I.E.5, I.F.1, I.F.2; II.B.8, II.F.3, II.F.5.
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4 W. Anderson, SCSB
- I.B.I.3, II.J.3.2, IV.E.4 h'[Norberg, EMSB
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\\. Parsont, RHSB
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ROErst B. m;;og.g,
Fobert B. ino;"e, Director Office of Standards Development cc:
W. Dircks R. Purple Distribution:
Central File RECORD NOTE:
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'each iten for'w'h3 hels responsibth.
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R. B. Minogue Input reflects input received from SD branches.
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W. M. Morrison J. L. Milhoan E. C. Wenzinger J. L. MIlhoan
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Director J. N0rberg I:E!'ORANDUM FOR:
01 vision ofTatety tecnnology W. Anderson.
Office of Nuclear Reactor T,egulation FRGil:
Robert B. !iinogue, Director Office of Standards Development SUSJECT:
SAFETY RATIONALE FOR ACTION PLAN Cy tc :crandum of July 11, 1980, the EDO requested SD provide to you a safety rationale far all TP.I Action Plan ite=s for which we have lead responsibility (as identified in Table 1 of flDREG-0660). Attached are initial drafts W tte
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ces-*-=t._30'tmc,hes m r.s%1.e for the SD identified tasks. The initial drafts were developed in accordance with the instructions contained in the July 11,1950 =emorandu::.
Cognizant SD Dranch Chiefs for the attached Action Plan ite.ts are as follows:
E. Uenzinger, RSSB - I.A.l.4, I.A.2.6, I.A.3.3. I.A.4.2, I.D.4, I.E.5, I.F.1, I.F.2; II.B.8,
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II.F.5.
W. Anderson, SCSB -
I.B.1.3, II.J.3.2, IV.E.4 J. ?!orberg, E!'SB - II.E.3.5 E. Conti, EPSB - III. A.2.1 R. Alexander,0HSB - III.D.3.2 H. Parsont, PJiSB -
III.D.3.5, IV.H Robert B. ;;inogue Director Office of Standards Development DISTRIBUTION CENTRAL FILE cc:
W. Dircks SD ALPHA SD RD RSSB WR SD:RSSB JLMILH0AN76.~
TASK NO. N/A RSSB
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ENCLOSURE Item No.:
I.A.1.4
Title:
Long-Term Upgrading Description of Action SD will develop proposed changes to 10 CFR 50 for consideration by the Commission to effect appropriate changes concerning plant staffing,-including shift manning, control room presence, and working hours. When revising the regulations, the staff will consider increasing the size of the shift operator complement by requiring the presence of two reactor operators and one senior reactor operator in the control room at all times during normal operations.
Provisions for working tours and status checks of the plant by individual operators normally assigned to the control room will be considered.
The results of the study of operator licensing (RFP-NRR-80-117) and the study of utility management and technical resources (NRC-03-80-105) will be considered.
In addition, the comments of the ACRS in its letter of December 13, 1979, will be considered.
Personnel requirements determined by emergency preparedness considerations will also be considered (refer to Item III.A.2.2).
Bases for Action 1.
Complex transients in nuclear power plants place high demands on the operators in the control room.
Theobjectiveoftheactions described in this task is to increase the capability of the shift crews 1
in the control room to operate the facility in a safe and competent manner by assuring that a proper number of individuals with the proper qualifications and fitness are on shift at all times.
Consideration must be given to both shutdown and operating conditions, since maintenance and shutdown conditions can affect operation.
2.
The intent of the proposed action is to establish the required levels of staffing of licensed operators for both shutdown and operating conditions.
Additionally, the proposed action should designate where these licensed operators should be located, in both shutdown and operating conditions.
3.
The proposed action is a proposed rule change to 10 CFR 50, " Domestic Licensing of Production and Utilization Facilities," which contains staffing proposals for licensed operators at commercial nuclear power plants.
The proposed action does not address production facilities or research and test reactors.
This proposed rule provides requirements for staffing which exceed the requirements approved by the Commission which are contained in NUREG 0694, "TMI Related Requirements For New Operating Licenses," and, therefore, the staffing requirements considered presently acceptable for the safety of the public are met or exceeded in the pro-posed rule.
This proposed rule meets the intent of the proposed action through establishment of required levels of staffing.
4.
There are no alternative means to address the intent of this proposed action since present regulations, which are clearly the only effective 2
l
m legal means of enforcement, are inadequate for staffing of licensed operators at commercial nuclear power plants and, therefore, need revision.
5.
Other actions which relate to this proposed action are:
a.
Proposed change to 10 CFR 55, " Operator Licensing," by revising required education, experience and training of licensed operators.
b.
Proposed change to Regulatory Guide 1.8, " Personnel Selection and Training," through increased education, experience and training recommendations, as well as establishing the Shift Supervisor as a recognized position.
c.
Proposed change to Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)," through establishment of specific working hour limitations.
d.
Proposed Regulatory Guide on Operator Fitness (Task I. A.3.3), to pro-vide recommendations on screening and evaluation of candidates for licenses to help eliminate possible sabotage or unexpected events caused by poor performance resulting from stressors on individuals.
3
Item No.:
I.A.2.6
Title:
Long-term upgrading of training and qualifications.
i Description of Action SD will develop new regulations and regulatory guides for training and qualifications of reactor operators, senior operators, shift supervisors, auxiliary operators, technicians, and possibly other ope *ating personnel.
Bases for Action 1.
A common concitsion of every investigation of the accident at THI-2 has been that many factors contributed to the accident but the major contri-buting factor was the manner in which the plant was operated both before and during the accident. Most studies have recommended changes in the numbers, qualifications, and organization of nuclear power plant personnel. Collectively, the studies have called for a general upgrading of utility capabilities for handling routine plant operations and for coping with unusual or unexpected conditions.
Chapter 1 of the NRC Action Plan Developed as a Result of the TMI-2 Accident (NUREG-0660) describes actions intended to improve operational safety, an area that has not previously been given the same regulatory emphasis as nuclear power plant design.
An important part of operational safety is the level of qualifications of operations personnel, includ!ng their education, training, experience, and fitness. One objective of this Action Plan item is to increase the level of the education of senior operators and other operations personnel to assure that they have appropriate technical backgrounds.
4
Besides educational background, training and experience of the operators and senior operators of nuclear power plants are being increased to improve their knowledge of plant design, plant response, and procedures.
The proper reaction to challenges requires a thorough understanding of plant design and plant response to transients and accidents, as well as training in the diagnosis and reaction to these conditions.
2.
The intent of the proposed action is to improve the capability of personnel to understand and control complex transients an'd accidents, and improve the general capability of en operations organization to perform their duties.
3.
The Action Plan item accomplishes the intent through a phased approach.
First research was initiated concerning:
(1) the means to be employed for selection and training of nuclear power plant personnel and the degree of NRC involvement in the process; (2) the means to be employed to evaluate the effectiveness of training programs including who, by job description, should be licensed; (3) methods to be employed to assure continued competency of plant personnel, including NRC involvement; (4) methods for maintaining a highly motivated and dedicated work force; and (5) means for rapidly requalifying presently licensed operators to meet the proposed new requirements.
Second, NRC regulations (10 CFR Part 55) and Regulatory Guide 1.8 (Personnel Qualification and Training) are being revised to incorporate previously issued short term requirements, changes resulting from the 5
national standards effort, and results of additional staff review in the drea'of personnel qualifications and training.
In this regard, a number of short term requirements have been issued by the NRC staff pending accomplishment of the long-term changes.
For example, in order to provide people with additional technical capability on shift until the time that staffing and qualifications of shift personnel and the control room man-machine interface requirements are upgraded, operating staffs are being required to have on-shift a technical advisor with engineering expertise, training in details of design, functions, arrangement and operation of plant systems, and special training in plant dynamic response.
(See Task I.A.1.1. of NUREG 0660).
Other short term requirements concerning personnel qualification and training are described in NUREG 0660 and include the following tasks:
Immediate upgrading of operator's training I.A.2.1 and qualifications i
I.A.3.1 Revised scope and criteria of licensing examinations I.A.4.1 Initial simulator improvement I.B.l.2 Evaluation of organization and manage-ment improvements of near-term operating license applicants I.C.1 Short-term accident analysis and procedures revision (including appropriate training) 6
Shift and relief turnover I.C.2 I.C. 3,
Shift supervisor responsibilities Training during low power testing I.G.1 Training for mitigating core damage II.B.4 Identification of and recovery from II.F.2 conditions leading to inadequate core cooling IE Bulletins on Measures to Mitigate II.K.1 Small-Break LOCAs and Loss of Feedwater Accidents Commission Orders on Babcock and Wilcox II.K.2 Plants III.D.3.3 -
Implant Radiation Monitoring.
Task I. A.2.6 involves the qualification and training of all nuclear power plant personnel, thus the task is considered sufficient in scope. With implementation of the short term requirements described above, timing is sufficient for development of revised guides and regulations and completion of the other subtasks (NRC training workshops, IE inspection procedures for training programs and review of the contents of'the basic course in nuclear power fundamentals in operator training programs).
4.
With regard to alternative methods for accomplishment of long-term upgrading of personnel qualification and training, it does not appear there are viable alternatives which permit public participation in the guidance development process.
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5.
Other actions of NUREG 0660 that relate or interact with Task I.A.2.6 to effect an overall improvement in personnel qualifications and training will be considered in the guidance development process.
The principal actions of NUREG 0660 which are involved are as follows:
I.A.2.2 Training and qualifications of operating personnel I.A.2.3 Administration of training programs I.A.2.5 Plant drills I.A.2.7 Accreditation of training institutions I.A.3.2 Operator licensing program changes I.A.3.4 Licensing of additional operations personnel I.A.3.5 Establish statement of understanding with INP0 and DOE I.A.4.2 Long-term simulator improvement I.B.1.1 Organization and management long-term improvements I.C.9 Long-term program for upgrading of procedures III.A.2 Upgrade emergency preparedness.
8
Item No.:
I.A.3.3
Title:
Requirements for Operator Fitness Description of Action A regulatory approach will be developed for Commission consideration to provide assurance that applicants for operator and senior operator licenses are psychologically fit (stress and malevolence), and to prohibit licensing of persons with histories of drug and alcohol abuse or with histories of criminal backgrounds.
Studies, criteria development, public comment, criteria issuance, and implementation are involved.
Two studies of interest are already underway in SD:
(1) standards for psychological assessment of plant personnel and (2) behavioral observation program to assure continued reliability of employees.
Bases for Action 1.
Upgrade the requirements and procedures for nuclear power plants operator and supervisor licensing to assure that safe and competent operators and senior operators are in charge of the day-to-day operation of nuclear power plants.
Increase the requirements for initial issuance of licenses and for license renewals and provide closer NRC monitoring of licensed activities.
In this regard, operator evaluations are being conducted which should provide (a) the necessity of an operator fitness program and (b) the requirements which should be contained in that program.
Preliminary results of the study indicate that there is a need to establish an operator fitness program, but the contents of that program are still subject to debate.
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a' 2.
The intent of the proposed action, which is a proposed regulatory guide, is to establish an operator fitness program that is acceptable to the NRC staff.
3.
The proposed regulatory guide on operator fitness should establish an opera-tor fitness program that will provide indicators of the psychological fitness of individuals, both for hiring, and as a continuing evaluation.
This proposed action has not been started since too little data is available at the present time to establish effective operator fitness program require-ments.
4.
The alternative to achieve the intent is a regulation.
Using a regulation to establish an operator fitness program will not allow the flexibility desired in a program which has so little a data base, and which is still subject to controversy when deciding on which programs are the most effective.
5.
Other actions that relate to the proposed action are:
a.
A proposed revision to 10 CFR 50.54, " Domestic Licensing of Produc-tion and Utilization Facilities," regarding minimum staffing of licensed operators.
1 i
b.
A proposed revision to Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation)," which contains specific working hour limitation recommendations, l
These are related since the numbers of individuals and their working hours can directly affect various aspects of psychological well-being.
I 10
Item No.:
I.A.4.2
Title:
Long-Term Training Simulator Upgrade Description of Action Conduct research on training simulators in the areas of simulator capabil-ities, operator's capability to recognize and cope with an accident situation (safety-related operator action) and operator error rates.
Upgrade training simulator standard ANS 3.5 and develop regulatory guide to endorse ANS 3.5.
Review simulators for conformance to criteria.
Bases for Action 1.
All of the reviews of the Three Mile Island accident have recognized the need for additional operator training.
A large part of this training was to include training to respond to abnormal, emergency, or accident condi-tions.
Since this training may be impractical or dangerous to perform on the actual plant, a simulator appears to be the best alternative.
The l
action would help ensure that the simulators used to conduct this training have the proper capabilities.
The action would also help establish the most effective methods for using the simulator in the training program.
2.'
The intent of the action is to: a) assure that simulators that are used for training properly represent the actual plant response and b) make l
l training program improvements.
3.
The action accomplishes the intent by conducting research on simulator and operator capabilities, upgrading training simulator standards with the results of the research, and reviewing the simulators in use to ensure conformance to the standards.
The action is sufficient in scope to achieve its intent.
The timing is sufficient to allow development of 11
the regulatory guide and to allow time for the owners of the simulators to comply with the criteria.
4.
The alternative to using simulators for this training is the use of actual plants or specific training plants to conduct training. The loss of power poduction capability for the plants and the time and cost of building plants specifically for training would make the alternative not viable.
The training provided by the plants would still not equal that of the simulators since major accidents could not be carried out realistically.
5.
Other actions that relate or interact with tnis action to effect an overall improvement are:
1.A.2.1 Immediate upgrading of operator training and qualifications.
Simulators will have to be capable of performing any simulator training required.
I.A.2.2 Training and qualification of operations personnel.
Same as I.A.2.1.
I.A.2.6 Long-term upgrading training and qualifications.
Sames as I.A.2.1.
I.A.3.1 Revise scope and criteria for licensing examinations.
Simulators will have to be capable of performing desired evaluations for licensing exams.
I.A.3.2 Operator Licensing program changes.
Same as I.A.2.1 and I.A.3.1.
I.A.4.1 Initial simulator improvement.
Any initial improvements will need to be evaluated for carry over to long-term upgrade.
I.A.4.4 Feasibility study of NRC engineering computer.
Research on engineering simulator should be carried over to improce response of training simulator.
12
Item No.:
I.B.1.3
Title:
Loss of Safety Function Description of Action SD prepared a staff paper presenting the following options related to regula-tory actions to be taken in the event of human or procedural error leading to a complete loss of safety function:
1.
Immediate shutdown if second such event in a two yea? 9eriod.
NRC approval required to resume operations.
Rule change involved.
2.
Use existing enforcement options (citations, fines, snutdowns).
No rule change.
3.
Use nonfiscal approaches such as point system, license probations and revocations.
No rule change.
Bases for Action 1.
The TMI-2 Lessons Learned Task Force (LLTF) Report (Short Term)
NUREG-0578 expressed concern over the frequency and consequences of human and/or procedural errors which could result in the loss of safety system operability.
SD was requested to prepare a rule change on this subject.
l IE objections to this approach became the second option put forward in the paper.
Task leader concerns with both options became the third option.
All three options address enforcement approaches toward errors of this type.
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2.
The purpose of this action is to reduce the incidence of preventable losses of safety system operability.
3.
Selection of an appropriate enforcement option would act as an adequate deterrent against such operational violations.
4.
This action contains its own alternative means with the three options pre-sented.
Selection of an appropriate option is a policy consideration for the Commission.
One of these three alternatives (option 2) has the poten-tial for being implemented by a related task (see #5), and this would eliminate the need for this task as a separate item.
5.
A related action is actually an outgrowth of the second option in this paper - development by IE of an overall NRC enforcement plan. With appro-priate consideration of relevant issues, this is actually a broader, and generally more desirable, approach.
Such a plan has already been discussed with the Commission and a revision is being prepared.
There are no related Task Action Plan Items.
14
Item No.:
I.D.4
Title:
Control Room Design Standard Description of Action SD will urge prompt revision of IEEE 566, " Recommended Practice for the Design of Display and Control Facilities for Central Control Roccs of Nuclear Power Generating Stations," and IEEE 567, " Design of Control Rooms for Nuclear Power Generating Stations." SD will issue a regulatory guide based on an evaluation of IEEE 566 and IEEE 567.
Bases for Action 1.
The various reviews of the Three Mile Island accident acknowledged that improvements in control room design would have at least helped the operators mitigate the effects of the accident if not prevented the accident.
A regulatory guide consisting of long-term corrective action requirements for existing control room designs and of requirements for future control room designs is necessary to ensure the control rooms are adequately improved.
l 2.
The purpose of the action is to propose a method that is acceptable to the NRC staff for complying with Criterion 19 of. Appendix A to 10 CFR Part 50.
3.
The action will accomplish its intent by incorporating the lessons learned from TMI-2 including Human Factors Considerations into the regulatory guide.
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4.
An alternative to this action is to develop a regulation instead of a regula-tory guide.
Developing a regulation in this area is not recommended at this time because of the different types of control rooms, and the flexibility needed to incorporate the improvements.
5.
Other actions that relate or interact with this action are:
I.D.1 Control room design review (information gained can be used in the standard),
I.D.2 Plant safety parameter display console (possible incorporation into the standard),
I.D.5 Improved control room instrumentation research (use research information in the standard).
4 16
Item No.:
I.S.5
Title:
Nuclear Plant Reliability Data System (NPRDS)
Description of Action NPRDS is a reliability oriented data collection and reporting system for selected components and systems related to the safety of nuclear power plants.
Periodic reports containing failure statistics are issued.
Licensee participation is voluntary and consequently inadequate.
An advanct notice of proposed rulemaxing to make participation in the NPRDS mandatory has been issued for public comment.
Bases for Action 1.
On April 29, 1977, a week after President Carter's energy message to a joint session of Congress, The National Energy Plan was published recommend-ing that the NRC make mandatory the present voluntary reporting of minor mishaps and component failures at operating nuclear power plants.
The Plan suggested that mandatory participation would enable industry and the NRC to develop a more reliable data base needed to improve reactor design, construction and operating practice.'
In working toward implementing of the National Energy Plan recommedation, several important issues have arisen which center around one of the Commission's own data reporting systems, the Licensee Event Reports (LERs),
and how the LER and NPRDS systems should be structured to complement each other and avoid duplication.
Furthermore, staff efforts have been guided by the basic re'quirement that data gathering must be shown to be necessary, not merely useful or interesting.
Coincident with NRC NPRDS activities, 17
l the Gen.ral Accounting Office (GAO) has reviewed NRC data gathering i
activities concerning unscheduled events at commercial nuclear facilities.
In a report issued in late January 1979, the GA0 concluded that it was unlikely the NRC could justify mandatory NPRDS participation when factors such as additional industry costs, limited expected safety benefits,'and duplication of NRC's LER system were considered.
However, the GAO believed that a full examination of the issue was warranted and suggested that the issue be decided using rulemaking procedures.
The GA0 concluded that rulemaking would provide the nuclear industry and the public, as well as the NRC staff, the opportunity to get their views on record and would better insure that all of these views are properly considered by the NRC.
GAO believed this to be particularly important since the GA0 believes the reliability data system has been developed and operated primarily by industry for industry's benefit.
2.
Following an April 19, 1979 Commission briefing on operational safety data gathering and analysis, the Commission concurred with the January 1979 GAO recommendation that rulemaking be used to decide the question of whether or not to make NPRDS reporting mandatory. Accordingly, an advance notice of proposed rulemaking (ANPR) was prepared, presented to (SECY 79-604),
and approved by the Commission and published in the Federal Register on January 30, 1980 (45 FR 6793).
3.
To date, 44 public comment letters have been received in response to the ANPR.
The NRC staff is currently preparing an analysis of those comments and a recommendation to the Commission on whether to make the NPRDS l
18
mandatory.
Offices providing direct support for this activity include, SD, AE00, MPA, NRR, IE, RES and ELD.
The most important factor for the NRC staff to decide whether to make a recommendation for a mandatory NPRDS is whether there is a well defined need in the NRC for the NPRDS informa-tion as opptsed to the information being merely useful or interesting.
Other factors in the NRC staffs' consideration of the matter include, but are not limited to, the status of other agency data gathering systems, most notably the LER System.
4.
Alternative means to achieve the intent have yet to be identified, but will be developed as the 44 public letters are reviewed and as recommenda-tions to the Commission are prepared.
5.
The need to upgrade the LER system, including revising Regulatory Guide 1.16, relates and interacts with this action item.
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l l
I 19 i
Item No.:
I. F.1
Title:
Expand QA List Descriotion of Action The NRC will develop criteria for licensees to expand their listing of equipment for which a quality assurance program is required to include equipment which is important to safety but has not been included in such listings in the past.
These criteria will be promulgated by means of both an amendment to 10 CFR 50, Appendix A and a Regulatory Guide, or guides.
Bases for Action 1.
Several systems which were proven to be important to safety during the Three Mile Island Unit 2 accident were not previously included on the list of equipment for which a quality assurance program is required (i.e.,
the Q list) and, therefore, were not designed, constructed or maintained in a manner commensurate with their important to safety.
Appendix A,
" General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 establishes requirements for structures, systems, and components impor-tant to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.
Criterion 1 of these general design criteria, " Quality Standards and Records," requires that a quality assurance program be established and implemented in order to provide ade-quate assurance that those structures, systems, and components will satis-factorily perform their safety functions.
Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Par.t 50 establishes quality assurance 20
requirements for the design, construction, and operation of certain struc-tures, systems and components; namely those that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. When these quality assurance criteria were published for comment on April 17, 1969, (34 FR 6599), the Statement of Considerations noted that they would supplement Criterion 1 of the General Design Criteria.
2.
Although it was recognized in developing both Appendix A and Appendix B that some structures, systems and components are more important to safety than others and, therefore, that it is not appropriate that all structures, systems and components which perform a safety function have the same quality assurance measures applied, the practice during design and construction of nuclear powr plants in many cases is to apply the same quality assurance requirements to all structures, systems, or components to which Appendix B applies.
This was never the intent of either Appendix A or Appendix B.
A graded approach to application of quality assurance requirements is l
specified in Critation 1 of Appendix A which states, " Structures, systems l
and components important to safety shall be designed, fabricated, erected and tested to quality standards commensurate with the safety functions to be performed," (emphasis added) and in Criterion II of Appendix B which states, "The quality assurance program shall provide control over activities affecting the quality of the identified structures, systems, and components, to an extent consistent with their imoortance to safety."
(emphasis added)
Despite the specification of a graded approach, the tendency has been to impose a " full blown" quality assurance program for some structures, systems, 21 l
and components important to safety while requiring no special quality assur-ance controls for others.
This has resulted in a more narrow definition of the structures, systems, and components that come under the quality assurance program requirements of Appendix B than was intended.
3.
As the first step in this effort to better define the ' equipment for which quality assurance requirements are to be specified, a clarifying amendment is being developed to Criterion 1 of Appendix A.
This proposed amendment would specifically state that the criteria for the quality assurance pro-gram required by the Appendix A, " General Design Criteria for Nuclear Power Plants," Criterion 1, are those criteria contained in Appendix 8.
Addi-tionally, it is the staff's intent to develop more detailed guidance, in Regulatory Guides, as necessary for determining both the specific equipment that is to be included on the list for which quality assurance requirements need to be invoked, and the extent of quality assurance requirements to be applied to specific structures, systems and components which are impor-tant to safety.
Both the a*.nndment to Appendix A and the additional guid-ance developed will be used to evaluate the quality assurance programs of both applicants and holders of construction permits and operating licenses for nuclear power plants.
4.
Other tasks that relate to this item are I.F.2 (Develop more Detailed QA criteria) to develop more detailed criteria for various aspects of the QA program and tasks II.C.1.1 and II.C.1.2 (Interim Reliability Evaluation Program) and II.C.1.3 (Systems interaction) which relate to defining the importance to safety of various types of plant equipment.
l' 22
o Item No.: I.F.2
Title:
Develop More Detailed Quality Assurance Criteria Description of Action The NRC will develop more definitive criteria for the various aspects of quality assurance programs for design, construction and operation of nuclear power plants.
Typical items to be considered under this task include more definitive qualifica-tions for cuality assurance personnel, definitive criteria for inclusion of quality assurance personnel in design, construction and operational phase activi-ties, criteria for determining quality assurance requirements for specific classes of equipment, criteria for adequate sizing of the quality assurance staff and clarifying the organizational reporting levels for the quality assurance organi-zation.
Bases for Action 1.
A recurring finding of the various groups that have analyzed the Three Mile Island Unit 2 accident has been that the role of the quality assurance function in plant activities has been inadequate.
This has resulted to some extent from application of a too narrow definition of which plant activities should come under quality assurance program requirements, and separate action in being taken to broaden the application of quality assur-ance requirements to various nuclear power plant systems (see item I.F.1).
2.
The proposed action is to develop criteria, in either regulations or regula-tory guides,'that will strengthen specific aspects of the quality assurance programs (as discussed in the description section above) that have been shown to be deficient.
23
3.
Although this item as a whole has been deferred until FY 1982, attention is being given to specific portions of this item, including qualification of quality assurance personnel and quality assurance organizational reporting requirements in response to specific identified needs concerning quality assurance problems such as those occurring during the construction of both the Marble Hill and South Texas projects.
4.
This task is closely related to item I.F.1 (expand QA list) in that a part of this task is to define quality assurance requirements for specific c1:sses of equipment.
24
Item No.:
II.F.3
Title:
Instrumentation for Monitoring Accident Conditions (Regulatory Guide 1.97)
Description of Action Revise Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," to include the lessons learned from TMI-2 as they pertain to accident monitoring, including the severity of. environmental conditions and the necessity for expanded ranges on certain instruments.
The revisions also considers the requirements for instrumentation relating to the Technical Support Center, the Safety Parameter Display console, the Emergency Operations Facility, the Nuclear Data Link, and control room upgrading.
Bases for Action 1.
The staff and the ACRS have for some years emphasized the need for special features and instruments for monitoring the course of an accident so that the operator is informed to take necessary mitigating actions.
Regulatory Guide 1.97 was issued in August 1977, containing a provision that four variables (specifically, containment pressure, reactor coolant pressure, radiation level inside containment, and plant radioactivity release rate) be measured with instruments that have expanded ranges and that are quali-fied to withstand a higher level of environmental conditions than that which is expected for design basis accident conditions.
Industry, in general, did not implement Regulatory Guide 1.97, because, in their expressed opinion, there was nothing in the regulations that required them 25
to consider anything beyond what was defined as Class 8 conditions.
How-ever, the climate for acceptance of instruments with expanded ranges has changed. The accident at TMI-2 demonstrated that conditions can arise that are more severe than those postulated for Class 8 conditions.
For instance, shortly after the start of the accident at THI-2, a 28 psig pres-sure spike occurred in containment which was a greater pressure than had
^
been postulated.
Additionally, the radioactivity effluent monitors and the core exit temperature monitors went off scale.
The accident at TMI-2 demonstrated further that there were some variables that should be displayed to the operator on a continuous basis that hereto-fore were not considered for such display.
Continuous indication of hydro-gen concentration inside containment and containment water level are two such measurements.
2.
The intent of the revision to Regulatory Guide 1.97, is to provide a systematic approach to determining variables that should be monitored for accident conditions, including postulated design basis accidents and un-postulated accidents. The revision also includes a list of variables that are considered to be a minimum set.
It is the responsibility of the licensee to complete the list of accident monitoring variables as applic-able to each individual plant.
3.
The implementation of certain key variables to be monitored, as defined by the lessons learned task force, has been imposed by NRR 1.etters to licensees and is to be installed in operating plants by January 1,1981.
1 A
26
The balance of accident monitoring instrumentation requirements are to be completed by June 1983.
Safety aspects and the difficulties of procuring and installing additions or modifications to in place instrumentation have been considered in establishing the schedule.
4.
There are no alternatives to monitoring variables in the plant to keep the operator informed as to plant status so that the operator can take
.necessary action to protect the health and safety of the public.
5.
Concurrent with the development of a revision to Regulatory Guide 1.97, a voluntary group from industry was formed to develop a consensus standard for accident monitoring instrumentation thus providing a broader base for establishing variables that should be monitored in the event of an accident.
27
Item No.:
II.B.8
Title:
Rulemaking Proceeding on Degraded-Core Accidents Description of Action Advance notice of proposed rulemaking will be published to provide the regulated industry and the public an opportunity to advise on the content of a regulation which will discuss the need for nuclear power plant designs to be evaluated over a range of degraded core cooling events with resulting core damage and the need for design improvements to cope with such events.
Bases for Action 1.
The accident at Three Mile Island (TMI) identified the need to reexamine our regulations and regulatory guides with respect to degraded cooling during severe accidents. This action comprises two separate portions.
One portion involves an Interim Rule concerning near-term actions on hydro-gen control and specific design and other requirements to mitigate the consequences of degraded-core accidents.
The Interim Rule covers specific recommendations, such as inerting of Mark I and II BWRs, that the staff believes to be of such safety significance that they should be codified by regulation as soon as possible.
The Interim Rule scheduled to be sent to the Commission for approval by August 1980.
2.
The other portion action identified in Task II.B.8 involves a long-term rulemaking which is to determine to what extent, if any, reactor plant designs and safety analyses should consider reactor accidents beyond those considered in the current design basis accident approach.
This includes a range of loss-of-core-cooling, core damage, and core melting events both 28
inside and outside historical design envelopes.
Furthermore, it is the intent of long-term rulemaking to' require more coherent consideration of this range of core damage events in the design of both normal operating systems and engineered safety features.
3.
As a first step in the long-term rulemaking, the staff is recommending that an advance notice of proposed rulemaking be issued to provide the regulated industry and the public an opportunity to advise the NRC on the cont'ent of a regulation winich would require improvements to cope with l
degraded core cooling and with accidents not covered adequately by tradi-tional design envelopes. With this input, the long-term rulemaking can better address the objectives of a regulation, the design and operational improvements being considered, and the cost of such design improvements compared to expected benefits.
The recommended advance notice of rulemaking contains questions concerning safety analysis, controlled filtered venting, hydrogen combustion control, core retention systems, inerting, and alternative design and operational improvements aimed toward coping with degraded cooling and melted cores.
These questions are expected to elicit detailed additional information concerning the foregoing key points to guide development of the long-term rule on degraded cooling.
i 4.
Alternative means to achieve the intent have not been considered since l
l this is te only practical means to achieve the desired results.
29
5.
There are a number of other rulemaking and regulatory actions that are closely related to degraded cooling rulemaking.
Some of these actions have been identified in the Commission Statement of Interim Policy on Nuclear Power Plant Accident Considerations under NEPA, published June 13, 1980, as well as in the Action Plan under items II. A, " Site Evaluation of Existing Facilities and Reformulation of the Siting Policy;" II.C, "Reli-ability Engineering and Risk Assessment;" and III. A.2, " Improving Licensee Emergency Preparedness - Long Term " These related rulemakings consists of:
(a) An advance notice of rulemaking concerning reactor siting which has been approved by the Commission for publication in the Federal Register.
Scheduled to Commission as proposed rule by April 1981.
(b) A minimum set of engineered safety features that will be required of all new plants as identified in the Advance Notice of Rulemaking for Reactor Siting.
Scheduled to Commission as proposed rule by April 1981.
(c) Upgrading of emergency preparedness for nuclear power plants.
Expected to be issued as an effective rule in July 1980.
(d) A proposed rule on Alternative Site Reviews issued on April 9, 1980.
30
Item No.:
II.E.3.5
Title:
Decay Heat Removal Description of Action II.E.3.5 Regulatory Guide 50 will issue for comment Revision 1 of Regulatory Guide 1.139, " Guidance for Residual Heat Removal to Achieve and Maintain Cold Shutdown," which includes changes to upgrade the residual heat removal (RHR) system to one that is important to safety and to reflect the impact of TMI-2 (e.g., the effect of highly radioactive source on system functional requirements, leakage, etc.).
This guide revision goes beyond the specific action described in Task II.E.3.5 of NUREG-0660, May 1980.
The specific action in NUREG-0660 includes only requirements for reaching cold shutdown using equipment that is important to safety and defers addressing guidance regarding a highly radioactive source pending the outcome of the interim and final rulemaking on degraded or melted core conditions (see Task II.B.8).
Bases for Action 1.
In nuclear power plant operation, experience shows that there have been and will continue to be accidental events that require long-term cooling of the reactor' system and bring the reactor to cold shutdown condition for inspection and repairs.
Currently, some systems, structures, and components required to perform this function are not designed as important to safety, i.e., safety grade. The Three Mile Island accident has rein-forced the need for a safe and reliable method to achieve cold shutdown, particularly under degraded core conditions.
31
2.
Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Removal System," provides the currcnt NRR position on RHR systems.
Proposed Revision 1 to Regulatory Guide 1.139 contains the basic regula-tory position of the branch technical position plus significant expansion to address the lessons learned from TMI.
Proposed Revision 1 to Regula-tory Guide 1.139 provides explicit guidance to address the assumption of high radioactivity in the reactor coolant and containment building environ-ment.
3.
One of the proposed actions in the guide recommends that systems and com-ponents necessary to achieve a cold shutdown in a power plant be designed as structures, systems, and components important to safety and that opera-tional procedures to perform this function be developed.
Current BWR designs meet the majority of these conditions, but current PWR designs do not; therefore, the following discussion will be limited to PWRs.
The proposed action addresses the following:
a.
Chemical and Volume Control System (CVCS) b.
Auxiliary Feedwater System (AFWS) and steam relief valves c.
Pressurizer and pressure control system (including pressurizer heaters and pressurizer-level indication) and auxiliary pressurizer spray, or pressurizer power-operated relief valves d.
Residual Heat Removal System (RHRS) e.
Operational procedures 32
Following the TMI-2 accident, cooldown using the RHR system was not attempted because of concern for possible leakage of highly radioactive primary coolant outside the containment.
Another action in the proposed revisiion of the guide addresses the requirements for bringing the reactor from normal operating conditions to cold shutdown even under the condi-tions of high radioactivity in the reactor coolant or containment building environment.
The proposed revision to Regulatory Guide 1.139 will be sent out for comment to obtain public response to assist in the development of the final guidance s
4&5. An alternative is to issue an effective guide with only requirements for safety grade equipment and to defer other requirements until rulemaking on degraded or melted core conditions is complete (i.e., the action currently in II.E.3.5 of NUREG-0660).
The reason for pursuing the current action of also including consideration of high radioactivity in Revision 1 of Regulatory Guide 1.139 and sending the guide out for comment is that the time required for rulemaking could be very long and it is desirable to get industry imput in this area in a more timely manner.
Additionally, l
the staff requirements for RHR systems, which are important to safety, are currently in effect via Branch Technical Position RSB 5-1.
The guide j
will again be revised to reflect the current NRC staff position whenever the rulemaking on degraded or melted core conditions is complete.
l 33
Item No.:
II.F.5
Title:
Classification of Instrumentation, Control and Electrical Equipment Description of Action IEEE, in conjunction with NRC, will prepare a standard to provide a classifica-tion approach for instrumentation, control and electrical equipment.
Bases for Action 1.
The current practice in the design and licensing of Nuclear Power Plants includes assigning electrical systems to either of two broad categories:
Class 1E or non-1E.
IEEE Standard 380-1975, " Definitions of Terms used in IEEE Standards on Nuclear Power Stations," defines Class 1E as:
"The safety classification of the electrical equipment and systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the environment."
Traditionally, reactor plant designers have judged equipment falling within the definition of Class 1E as " safety grade" while considering most instru-m?ntation, control, and electrical equipment falling outside the definition of Class 1E as non-safety grade.
Furthermore, in reviewing reactor plant designs using the " design basis accident" approach, the NRC does not review all structures, systems, and components but reviews, in varying levels of detail, only those considered
" safety grade" by the applicant submitting a Safety Analysis Report.
Items considered by the applicant to be outside the scope of design basis accident analyses are generally not considered to be " safety grade" and are not 34
reviewed by the NRC to see whether they will perform as intended or meet various dependability criteria. This method of classification is based on the notion that things credited in the analysis of a design basis event or specified in the regulations are important to safety and, thus, are
" safety grade" while all else is "non-safety grade." Non-safety grade items do not receive continuing regulatory supervision or surveillance to see that they are properly maintained or that their design is not changed in some way that might interact negatively with other systems.
- Instead, these items simply receive what attention may be dictated by routine indus-trial codes and by desires to enhance plant availability.
2.
The accident at Three Mile Island has shown the need to reexamine these historical approaches to instrumentation, control, and electrical system design, and design review.
For example, the October 1979 Report of the President's Commission on the Accident at Three Mile Island stated that:
...there is a sharp delineation between those components in systems that are ' safety-related' and those that are not.
Strict reviews and requirements apply to the former; the latter are exempt from most requirements--even.though they can have an effect on the safety of the plant. We feel that this sharp either/or definition is inappropriate.
Instead, there should be a system of priorities as to how significant various com-ponents and systems are for the overall safety of the plant."
Similarly, the January 1980 report, Three Mile Island, A Report to the Commissioners and to the Public states:
"The current classification of systems and equipment into ' safety related' and 'non-safety related' is especially unsatisfactory."
The reports goes on to state:
"The process is not good enough to pinpoint many important design weaknesses or to address all the relevant design issues.
Some important accidents are out-side or are not adequately assessed within the ' design envelope';
key systems are not ' safety related;' and integration of human factors into the design review is grossly inadequate."
35
o The purpose of the new standard now under development is to present a uniform classification approach for determining the applicability of design criteria and design requirements for nuclear power generating station equipment, based on the level of the equipment's importance to safety.
3.
The scope of the new standard is broad, setting forth criteria for deter-mining the level of importance to safety of the instrumentation, control and electrical portions of nuclear power generating station equipment and focusing primarily on the equipment not covered by IEEE Standard 603.
" Trial-Use Standard Criteria for Safety Systems for Nuclear Power Generating Stations," August 15, 1977.
This new standard, in concert with a new regulatory guide, will provide a method for determining the relative import'ance to safety of instrumenta-tion, control, and electrical systems, and will provide a method for deter-mining the applicability of design requirements to these systems.
These documents will cover, in a coherent fashion, the whole range of instrumenta-tion, control, and electrical systems important to safety, including reactor protection equipment and engineered safety features presently addressed by the Class 1E practice.
4.
Alternative means to achieve the intent have been considered.
- However, interpretation and application of the requirements established for nuclear power generating station reactor protection systems and engineered safety features to the design and operation of systems that are not part of the reactor protection system or engineered safeture features.is very complex.
36
It does not appear to lend itself to a simple classification system which would clearly and simply define system and component requirements for a designer. Therefore, the new standard will provide criteria for determining the degree of tpplicability of the requirements of other standards. Using this approach, the end product of the standard would not necessarily be a system of discrete classes for systems or equipment.
Based on this approach, the IEEE/NRC working group intends to develop a standard which provides criteria for determining the level of importance to safety of systems and a method for determining requirements for systems as related to their importance to safety.
5.
A related action is a proposed amendment (under development) to 10 CFR 50 Appendix A.
See Task I.F.1.
l 37
Item No.:
II.J.3.2
Title:
" Management for Design and Construction of Nuclear Power Plants" Descriotion of Action SD will issue a regulatory guide that codifies the requirements for technical resources and controls during the design, construction, and modification phases.
Bases for Action 1.
Reason - In the aftermath of The Three Mile Island nuclear power plant, studies and investigations have been conducted by the industry, the NRC, and others to search for the basic causes that led to the accident and to come out with recommendations, which, when appropriately implemented, would~
prevent such occurrences in the future.
Collectively, these studies have called for a general upgrading of utility capabilities for handling routine plant operations and accident conditions. The proposed item is one item inTMIActionPlanwhichwouldensurethattheapplicant/'licenseepossesses adequate technical and management resources to oversee the design and con-struction (or modifications) of the plant.
2.
Intent - The criteria, when properly implemented, would ensure that the utility-owner possesses the first hand knowledge regarding the structures, systems and components necessary for the safe operation of the plant.
This knowledge and familiarity with the plant systems and components will enable the utility to take appropriate actions without delay to prevent a major accident due to failure or malfunction of a system or a component.
38
O 3.
Action Accomplishment and Scheduling - The criteria. developed in this action will require an applicant-licensee to develop adequate resources to oversee the design and construction of the plant (including modifications to operating plant) by considering the number of people to be used as well as the areas of expertise, competency, and scope of work to be performed.
The criteria will also delineate the degree of management and technical control to be exercised by the utility during design and construction, including the preparation and implementation of procedures necessary to guide the effort.
This effort (i.e., to develop regulatory guide describing criteria) is a high priority item.
The sequence and timing established for this action are tight, but practically feasible.
4.
Alternatives to the Action - A clear alternative to not having these criteria is to have the utilities oversee the design and construction in a usual manner.
The NRC staff believes this is not acceptable.
During the imple-mentation of these criteria the subject of degree of control needed by applicants-licensees during design and construction will be discussed, alternatives scrutinized and the necessary refinements will be made. The procedural alternatives to developing the action (i.e., developing a regulatory guide) would be to develop a rule or endorse an ASME developed "0wners Certification Program." However, at this time, publishing a regulatory guide is considered to be the most efficient approach.
5.
Interaction with Other TMI Actions - Task I.B.1 " Management for Operations" is closely related to the Task II.J.3, " Management for Design and Construc-tion." The staff and resources developed by a utility as a result of the criteria in Task II.J.3 will be quite useful in developing capabilities 39
O required for Management for Operation of nuclear power plant.
Thus, in combination, the capabilities developed by utilities to meet these criteria will enable him/her to respond to various malfunctions and failures expeditiously and efficiently which in turn would eliminate any chances for occurrence of a major accident.
40 i
Item No.:
III. A.2.1
Title:
Improving Licensee Emergency Preparedness - Long Term Description of Action Modify regulations to upgrade the requirements for emergency preparedness at nuclear power plants.
Bases for Action 1.
The Commission determined as a policy matter to upgrade requirements for emergency preparedness as a condition for granting a license to operate a nuclear power plant. Preplanning and demonstrated capability to provide immediate notification and information on protective actions enhances the capability to minimize public exposure received following a nuclear power plant accident.
2.
See Item 1.
3.
A final rule was affirmed for issuance in the Federal Register by the Commission on July 23, 1980.
4.
Not applicable.
5.
Revision of Regulatory Guide 1.101 incorporating emergency planning criteria in NUREG-0654 Revision of NUREG-0610 (provides definition of reactor accident scenarios)
Development of a Nuclear Data Link (III.A.3.4)
The Reactor Siting Rule, Degraded Core Rule, and 41
---r y,
-m
l Regulatory Guides on Realistic Meteorology for Accidents are closely related to the Emergency Planning Rule I
l l
f l
l l
1 l
)
42
Item No.:
III.D.3.2 Titl e:
Health Physics Improvements (10 CFR 20)
Descriotion of Actions 1.
Amendment of 10 CFR Part 20 to require that personnel dosimetry processing be done only by nationally certified processors meeting specific performance criteria.
2.
Issue a regulatory guide containing specifications for audible alarm dosimeters and criteria for their use.
The proposed action will provide guidance on the selection of reliable audible-alarm dosimeters and on the appropriate use of such dosimeters by nuclear power plant workers, and others, as a supplemental warning of exposure to radia tion. The guidance is intended to allow licensees to select reliable audible-alarm dosimeters and to provide dosimeter manufacturers with criteria they can use in designing their products.
3.
Develop standard performance criteria for radiation survey and monitoring instruments; contract for performance testing of off-the-shelf instruments to determine the feasibility of the criteria; and amend 10 CFR Part 20 to require that only those instruments meeting the criteria in the standard may i
be used at licensee facilities.
Bases for Actions 1.
The TMI accident focused attention on the inadequacies of personnel dosimetry, the need for reliable, accurate survey and monitoring instruments, and the risks associated with exposure to radiation, including occupational exposures.
The NRC's TMI Special Inquiry Group recomended (1) that occupational radiation l
protection, "which has always been secondary to reactor operations and reactor safety," be given higher priority; (2) that the NRC " require licensees to have adequate personnel dosimetry services"; and (3) that "in-plant and portable radiation monitoring instruments...be available at all nuclear power plants."
43
The President's Commission recommended improvement of methods of monitoring and surveillance, including epidemiologic surveillance, to monitor and determine the consequences of exposure to radiation of various population groups, including workers. The President's response to the recommendations of the President's Commission included direction to DOE to strengthen its program to develop technologies for reducing the radiation exposure of workers at nuclear power plants. The President's Commission also recommended that there be a sufficient supply of health-related instruments at nuclear power plants for both routine and emergency conditions.
The objective of Task III.D.3, " Worker Radiation Protection Improvement," is consistent both with the above-mentioned recomendations and the President's directive.
2a. Assessing, reducing and otherwise controlling the exposures of workers is critically dependent on consistently accurate measurement of such exposures by means of personnel dosimetry. Studies have established that an acceptable level of consistency and accuracy does not exist throughout the personnel dosimetry service field. Therefore, it is essential that the performance of many personnel dosimetry service organizations be improved.
- b. The inappropriate use of inadequately rugged and unreliable audible alarm dostmeters could cause unnecessary exposures by failing to provide a warning to the wearer with respect to unknown radiation levels or unexpectedly high radiation levels such as are likely to be encountered in cleanup operations such as those at TMI. Further, since such dosimeters are currently on the market, the availability of important information concerning their characteristics (including deficiencies) is important to their proper use and will help prevent exposures that might otherwise occur because of the improper use of the devices based on a lack of knowledge concerning these characteristics. The guide will discourage the use of unreliable dosimeters and the inappropriate use of the 44 dosimeters such as in those cases where rolatively high, but unprodictable radiation levels may occur.
- c. While it is anticipated that more; attention will be given to the provision of adequate supplies of survey and monitoring instruments at nuclear power plants as a result of the TMI accident, experience has established that there are wide variations in reliability and accuracy among the survey and monitoring instruments currently on the market. Clearly, a standard is needed as a basis for the manufacturers of such instruments to work from and as a criterion by which users of such instruments may judge their adequacy.
3a. The proposed action with respect to personnel dosimeter services will affect all service suppliers of NRC licensees.
Interagency coordination and cooperation was initiated very early in NRC efforts to upgrade personnel dosimetry programs with the informal formation of the Interagency Policy Committee on Personnel Dosimetry Performance.
Represented on this committee are:
the Bureau of Radiological Health (HEW), the Department of Defense, the Department of Energy, the Environmental Protection Agency, the National Bureau of Standards, the Occupational Safety and Health Administration (00L), the Conference of Radiation Control Program Directors (States), and the NRC. The representatives of the agencies on the committee consider the NRC to be the lead agency for the overall problem, and they plan to recommend to their various agencies that regulations or intra-agency rules similar to the NRC regulations be adopted.
- b. The presentation of presently available audible alarm dosimeters' characteristics in conjunction with standard specifications to appear in the guide should result 1
in more intelligent selection of dosimeters by users and the production of more suitable dosimeters by the manufacturers. The guide on audible alarm dosimeters has been issued in draft form and should be available in final form as the cleanup activities intensify.
- c. The effectiveness of survey and monitoring instruments should be greatly improved as a result of the proposed regulatory standard by helping to assure that those 45 l
instruments selected for each nuclear power plant meet certain minimum standards of reliability and accuracy. The proposed action would affect all manufacturers and suppliers of survey and monitoring instruments who provide instruments to NRC licensees since such licensees would be required to use instruments meeting the criteria of the rule. These same manufacturers and suppliers provide in-struments to those portions of the nuclear industry not subject to NRC control such as certain DOE and D0D activities and activities in Agreement States. There-fore, NRC action in this area will likely also effect imprcvement in such instruments provided throughout the nuclear industry.
4a. An alternative approach to improvement of the consistency and accuracy of personnel dosimetry services would consist of waiting for the industry to adopt and implement, consistently, performance standards resulting in the desired level of quality. However, the deficiencies outstanding in the personnel dosimetry services field have been with us for decades and are not likely to be overcome without the impetus of a rule change to stimulate appropriate action throughout the industry.
- b. With respect to audible alarm dosimeters, the NRC could, as an alternative course, wait for the industry to develop, adopt and implement a consensus standard. This is not considered a practical course to follow because of the rather lengthy delays inherent in such a standards development and implementation process and the current need for guidance.
- c. With respect to survey and monitoring instrument improvement, an alternative approach would consist of allowing industry to develop, adopt and implement a consensus standard for survey and monitoring instruments. However, problems of instrument unrealiability and inaccuracy have been with us for years and are not likely to be overcome without the impetus of a rule change to stimulate appro-priate action throughout the industry.
5.
None.
46
Item No.:
III.D.3.5
Title:
Radiation Worker Exposure Data Base Description of Action Continuation of efforts to improve and expand the data base on industry employees to facilitate possible future epidemiological studies on worker health.
Bases for Action 1.
To improve nuclear power plant worker radiation protection it is necessary to develop a data base which will allow determination of the adequacy of current radiation protection standards.
Presently, there is no standard format across the industry for collection and centralization of sufficient data for potential epidemiologic studies of worker health.
2.
The intent of this action is to develop a format for data pertinent to eventual epidemiologic studies, and to investigate methods for using these data in such studies so as to provide a better empirical basis for the development of radiation protection standards.
3.
This action accomplishes its intent by examining the existing data now collected by the industry, by interacting with other Federal agencies concerned with worker health, and by utilizing the results of studies now underway to develop data requirements for external and internal doses, medical radiation exposures, health data, and exposure to nonradioactive carcinogens, as well as data to facilitate long-term follow-up of workers.
47
~
~
4.
The alternative exists for other Federal agencies such as OSHA or HHS to perform this action. Along these lines, there do exist limited registries -
for Plutonium workers, Uranium workers and Uranium miners.
In addition, DOE has initiated a program to establish an extensive computerized data base on its employees and the employees of its contractors. NRC has already contacted DOE to initiate cooperation in this area. NRC, in cooperation with NIOSH, has been involved in the establishment of a radiation worker registry at TMI since April 1979, and a nuclear power industry-wide worker registry since shortly thereafter.
Furthermore, it is very likely that the NRC authorization bill for 1981 will contain a mandate for NRC to investigate and report on the status of health data on licensee employees. Outside of DOE, and the NRC/NIOSH effort however, no other Federal agency has initiated such broad activities for radiation workers.
In order to facilitate the establishment of a meaningful data base, it is necessary to have appropriate legislation. Otherwise it becomes a matter of dependence on the cooperation of licensees. Although possible for another agency to perform this action, it is more efficient for NRC to continue to complete the activities which it has initiated by itself, in cooperation with NIOSH, and under the mandate of the Congress.
5.
None.
48
,.3.
Item No.:
IV.E.4
Title:
Resolve Generic Issues by Rulemaking Description of Action To establish criteria to be utilized by the staff in identifying candidate safety issues and evaluating their suitability for rulemaking procedures.
Bases for Action 1.
The Kemeny Commission indicated the need for the NRC to improve upon prior performance in resolving generic and specific safety issues. They recommended that " issues that recur in many licensings should be resolved by rulemaking" in order to eliminate repetitive consideration of some issues in that process.
2.
The ultimate intent of this action is to use rulemaking as an instrument for improving the effectiveness of the licensing process.
3.
The action is quite specific and directed at meeting the intent of the Kemeny Commission recommendation within a time frame consistent with resolution of other action plan items which could be implemented through rulemaking.
4.
Alternative means of resolving generic issues is through license changes, orders or changes in regulatory guides. These latter approaches have not proven to be as enforceable as regulations nor have they eliminated repetitive consideration in licensing proceedings.
Implementation of the resolution of generic issues thru rulemaking has the potential for improving the present licensing procedures.
49
I-
.a>g 5.
Other actions which relate to this action are the development of plans for early identification, assessment, and resolution of safety issues so that implementation thru rulemaking can be accomplished.
1 e
i 4
i l
f -
I i
6 l
1 I
i i
i I
i 50
I
.=,h Item No.:
IV.H
Title:
NRC participation in the Radiation Policy Council Description of Action NRC participates in the Radiation Policy Council through its designated member, the Director, Office of Standards Development and NRC staff representatives on th,e Working Group and Task Forces.
Bases for Action In the President's response to the recommendations of the President's Commission on the Accident at Three Mile Island he announced initiatives to establish a Radiation Policy Council and an Interagency Radiation Research Committee (formally established by executive order on February 21,1980).
The response also stated that the President was taking additional actions, one of which was to request NRC to submit for review all actions affecting worker and public health and safety to the Radiation Policy Council.
In response to this request, NRC has been participating fully in the Radiation Policy Council proceedings.
51
.