ML19340B361
| ML19340B361 | |
| Person / Time | |
|---|---|
| Issue date: | 09/04/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19340B362 | List: |
| References | |
| TASK-OS, TASK-RS-902-4 REGGD-01.033, REGGD-1.033, NUDOCS 8010220168 | |
| Download: ML19340B361 (49) | |
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Proposed Rule (PR)
ACRS Minutes No.
Reg. Guide
/J G /. 23 Relates to Proposed Rule (PR)
Petition (Pati)
Relates to Reg. Guide Effective Rule (R !)
Relates to Petition (P.:Ji)
A?iSI Relates to Effective Rule [P:ij j
IAEA Federal Registc'r Motice SD Task No. X'S 90 2-4 NUREG Report Contract No.
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September 4, 1980 l
SECOND PROPOSED REVISION 3* TO REGULATORY GUIDE 1.33
_ QUALITY ASSURANCE PROGRAM REQUIREMENTS l
(OPERATI0fi)
A.
INTRODUCTION Appendix A, " General Design Criteria for Nuclear Power Plants," to 10 CFR
-Part 50, " Domestic Licensing of Production and dtilization Facilities," estab-lishes requirements for structures, systems, and components important to s:fety; 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
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implemented in order to provide adequate assurance that those structures, systems, and components will, satisfactorily perform their safety functions.
Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 establishes quality assurance require-ments-for the design, construction, and operation of certain structures,-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. ' The criteria for the quality assurance program required by the Appendix A, " General Design Criteria for Nuclear Power Plants," Criterion 1,
- The_ first proposed Revision 3 to Regulatory Guide 1.33 was issued for public comment in August 1979.
S'nce that date, a substantial amount of guidance concerning quality assurance has-been develcped through assessment lof the-l Three Mile Island Unit 2 accident by various organizations.
In addition, l
ANSI N18.7-1976/ANS 3.2 which is endorsed by the regulatory guide is undergoing i
extensive revision in an effort to provide more definitive quality assurance-l' program requirements.. As'a-result of the incorporation of additional guidance L
into the revisions to the ANSI standard and the regulatory guide,-proposed ~
Revision 3 to Regulatory Guide 1.33 is being issued'for a second public comment period;in order ~to obtain additional public input on the proposed regulatory guidance.
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are those criteria contained in Appendix B.
This regulatory guide describes a method acceptable to the NRC staff for complying with the Commission's regula-
- tions with regard to overall quality assurance program requirements for the operation phase of nuclear power plants.
'B.
DISCUSSION Subcommittee ANS-3, Reactor Operations, of the American Nuclear Society
}
Standards-Committee developed ANSI N18.7-1972, which contained criteria for administrative controls for nuclear power plants during operation.
This j
standard, along-with ANSI N45.2-1971, " Quality Assurance Program Requirements for Nuclear Power Plants," was endorsed by Regulatory Guide 1.33.
The dual endorsement was necessary in order for the guidance contained in the regulatory guide to be consistent with the requirements of Appendix B to 10 CFR Part 50; i
however, this dual endorsement caused some confusion among users. To clarify
- this situation, ANSI N18.7-1972 was revised so that a single standard would r
- define the general quality assurance program " requirements" for the operation phase. This revised standard was approved by the American National Standards Committee N18, Nuclear Design Criteria.
It was subsequently approved and designated ANSI N18.7-1976/ANS-3.2, " Administrative Controls and Quality 4
"1 Assurance for the Operationa' of Nuclear Power Plz by the American i
National Standards Institute..-ebruary 19, 1976.
Revision-I to Regulatory Guide 1.33 was issued for public comment in-1 January 1977.
Following the public comment period, the regulatory guide was further revised to reflect-public comments and additional' staff input. As a result, Revision 2 to Regulatory Guide 1.33 was issued in February 1978.
As a result of the assessment of the Three Mile Island - Unit 2 nuclear 1
power plant accident by the industry, the NRC, and others, improvements to the quality assurance program requirements for the operation of nuclear power. plants have been recommended. A number of studies and investigations that have been conducted have a significant impact on the guidance provided in this regulatory-i
- guide. ANSI N18.7/ANS 3.2 is undergoing extensive revision to incorporate the substantial amount 'of guidance that has been developed from the various studies and. investigations. One aspect of this revision to the standard is the inclusion ICopies may be obtained frow the American Nuclear Society, 555 North Kensington Avenue,- La Grange Park, Illinois 60525.
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of Appendix A, " Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors," of Regulatory Guide 1.33.
Other areas addressed in the draft revision to the standard are operator and shift supervisor authority and responsibilities, maximum working hours for operating personnel, control of access to the control room, additional equipment control requirements, and more explicit guidance to provide an improved quality assuranca program for the operational phase of a nuclear power plant.
Due to the length of time necessary for a draft revision to a national standard to be accepted and published, the NRC staff with the consent of the ANSI organization has decided to endorse Draft 5 (dated August 1980) of the revision to ANS 3.2 subject to the exceptions delineated in Section C of this guide. At that time when the revision to ANS 3.2 is published, an early revision to Regulatory Guide 1.33 will be initiated to endorse the published ANS 3.2 document subject to possible exceptions.
Addressing all of the completed, current, and future studies and investiga-tions of the TMI-2 accident would be impractical. However, the principal efforts that have been useful in developing the revisions to the subject national stand-l l
ard and regulatory guide will be discussed below:
1.
NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 l
Accident,u2 was developed to provide a comprehensive and integrated plan for j
the action now judged necessary by the NRC to correct or improve the regulation l
and operation of nuclear facilities based on the experience from the accident l
at TMI-2 and the official studies and investigations of the accident. The discrete scheduled tasks identified in the action p;an were developed from the rec.mmendations of organizations who investigated the accident. These organiza-tions include the Congress, the General Accounting Office, the President's Commission on the Accident at Three Mile Island, the NRC Special Inquiry Group, the NRC Advisory Committee on Reactor Safeguards (ACRS), the Lessons-Learned "The recommendations of the investigating groups are collected in NUREG-0660, "NRC Action Plan Developed as a Result of'the TMI-2 Accident." Copies of NUREG-0660 may be obtained from: GPO Sales Program, Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washing-ton, DC 20555, and National Technical Information Service, Springfield, VA 22161. NUREG-0660, in Appendix E, discusses the availability of the individual investigatory reports.
3
l Task Force and the Bulletins and Orders Task Force of the NRC Office of Nuclear Reactor Regul'ation, the Special Review Group of the-NRC Office of Inspection and Enforcement, the NRC staff Siting Task Force and Emergency Preparedness Task Force, and the NRC Offices of Standards Development and Nuclear Regulatory.Research. Therefore, the individual documents resulting from the investigations conducted by these organizations will not be addressed in this regulatory guide.
-Due to the broad scope of activities in'which quality assurance is involved, many of the tasks. described in NUREG-0660 will directly or indirectly affect the guidance provided in the ANS 3.2 standard and Regulatory Guide 1.33.
l 2.
In early 1980, Basic Energy Technolcgy Associates, Inc. (BETA) com-l pleted a study for the Office of Nuclear Reactor Regulation which outlined the results of a comparative review of current NRC requirements, licensed nuclear power plant practices and the Naval Nuclear Propulsion Program procedures for the selection, training and qualification of personnel involved in nuclear plant operation and maintenance. The results of the BETA study entitled,
" Power Plant Staffing," are documented in NUREG/CR-1280/ BETA-1033 cn which public comments were requested. The recommendations of this study are currently l
being considered for their applicability to Regulatory Guide 1.33.
Three l
recommendations have been used in providing improved guidance in this proposed revision to Regulatory Guide 1.33.
These recommendations involved (1) equipment i
or system failure reporting (Regulatory position 5), (2) substitutions of specified parts (Regulatory position 5), and (3) operator working hours j
(Regulatory position 9).
3.
Commission Information Paper SECY-80-2424 discusses the establish-ment of a group, referred to as the Indepandent Safety Engineering Group, that l
is independent of the plant staff, but is assigned onsite to perform independent
" Copies may be cotained from: GPO Sales Program, Division of Technical Informa-tion and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC l
20555, and National Technical Information Services, Springfield, VA 22161.
l 4 Document is available for inspection and copying for a fee in the NRC Public Document Room at 1717 H Street, Washington, DC 20555.
4 "S
reviews of' plant' operational activities and to provide a capability for evaluation of operating experiences at nuclear power plants. Task I.B.1.2 of NUREG-0660 discuss'es the requirement for the establishment of this group by each near-term operating license applicant. Appendix A to this proposed revision to Regulatory Guide 1.33 contains an excerpt from Commission Informa-tion Paper SECY-80-242 that' discusses'this group including Table 1 that lists the responsibilities of the various review groups.
4.
The NRC staff has'under development criteria for onsite and offsite organizations, both management a'nd technical, including the radiological pro-
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.tection organization, that will provide assurance of the safe operation of the 4
plant during normal'and abnormal conditions and the capaoility necessary to respond -to unusual: or unexpected situations. A contractor was selected (NRC-03-80-105, TEKNEKRON Research, Inc.) to assist in criteria development.
j TEKNEKRON submitted its final report entitled, " Utility Management and Techni-3 cal Resources,".to the NRC in May 1980.
This report is currently being
)
reviewed to determine its impact on the guidance provided in Regulatory
' Guide 1.33.
Task I.B.1.1, " Organization and Management of Long-Term Improve-j' ments," and Task I.B.1.2, " Evaluation of Organization and Management Improve-j
- ments of Near-Term Operating License Applicants," of NUREG-0660 describe the j
criteria development effort. As described in Task-I.8.1.2, near-term operating j-license applicants are being required to comply with the findings and require-3 ments generated in an interoffice NRC review of. licensee organization and management. -The review is based,-in part, on an NRC document entitled " Criteria j
for Utility Management and Technical Competence."4 A draft of this document l-was dated July,17, 1980, but.the-document is changing with use and experience
.in ongoing reviews. The document addresses the organization, resources, f'
training,.and qualifications of, plant staff, and management (both onsite and-
)
offsite) for routine' operations and accident conditions. As the final criteria l
aredevelopedf,RegulatoryGuide1.33willbereviewedtoensureconsistencyin the. applicable areas.
I 5.-
~In February:1980, IE Circular No.E 80-024 on Nuclear Power Plant Staff Work Hours was' issued..and discussed the fil effects that can result from extended i
personnel work hours. The'IE Circular recommends a maximum work hour schedu!e
.for;certain personnel. 'In addition,;a preposed rule change to 10 CFR 50.54 is currently.under development that will.establ.ish= requirements in this area.
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The draft ANS 3.2 standard presents a work time schedule that is generally consistant with the IE Circular No. 80-02.
A regulatory position (#9) is included in this proposed revision to Regulatory Guide 1.33 that expands the guidance provided in the draft ANS 3.2 standard.
6.
In January 1980, NUREG-0654/ FEMA-REP-1, " Criteria for Preparation l
and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,"3 was published for interim use and comment.
The purpose of this interim guidance and upgraded acceptance criteria is to provide a basis for NRC licensees, state and local governments to develop radiological emergency plans and improve emergency preparedness. The guidance is the product of the joint FEMA /NRC Steering Committee established to coor-dinate the agency's work in emergency preparedness associated with nuclear power l
plants.
It will be used by reviewers in determining the adequacy of state, local and nuclear power plant operator emergency plans and preparedness.
The guidance provided in the document will be published in final form after public comments are reviewed and resolved. A regulatory position (#1.d) of this proposed revision to Regulatory Guide 1.33 references NUREG-0654 for guidance in developing emergency plans. Table B-1, " Minimum Staffing Requirements for NRC Licensees for Nuclear Power Plant Emergencies," of NUREG-0654 discusses minimum capabilities and staffing on-shift and available within 30 minutes l
following the declaration of the emergency.
Section 3.4.3, " Technical Support l
for the On-Duty Operating Staff," of the draft ANS 3.2 standard requires that l
technical support personnel be available within two hours of the commencement of an accident. This propcsed revision to Regulatory Guide 1.33 references l
NUREG 0654 with respect to the availability of technical support personnel following the declaration of the emergency.
7.
In May 1980, NUREG/CR-1368/ SAND 80-7053, " Development of a Checklist for Evaluating Maintenance, Test and Calibration Procedures Used in Nuclear Power Plants,"3 and NUREG/CR-1369/ SAND 80-7054, " Procedures Evaluation Checklist for Maintenance, Test and Calibration Procedures,"3 were published as a result 1
of a study done by Human Performance Technologies, Inc. ' for the Division of Reactor Operations Inspection of the Office of Inspection and Enforcement.
NUREG/CR-1368 describes the process for. developing a checklist to be used by
_I&E inspectors during their evaluation of maintenance, test and calibration procedures.
NUREG/CR-1369 describes the' checklist and provides instruction 6
l
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h for use by I&E. inspectors in implementing the checklist. Analysis of these documents is currently' underway and Regalatory Guide 1.33 will be revised 'as improved guidance is developed.
8.
As a result of an NRC staff review of the operator licensing program a proposed rule change is currently under development that will amend 10 CFR j
Part 50 and.10 CFR Part 55. The proposed changes to Part 50 and 55 were forwarded to the ACR$ for review on May 14, 19804 and discussed with the ACRS t
Regulatory Activities Subcommittee on June 4,1980.
The purpose of the amend-ments to 10 CFR Part 55 is to specify requirements designed to improve operator performance in order to prevent accidents induced by the operator and to improve the operator's ability to mitigate an accident if it should occur.
The purpose l
of the amendment to 10 CFR Part 50 is to ensure an operator is familiar with the current plant condition prior to being allowed to manipulate 'the controls or supervise the manipulation of the controls of the facility.
Specifically l
Paragraph 50.54(r) is proposed to be added as follows:
" Administrative pro-
.cedures shall be developed by the licensee to provide assurance that an opera-tor or senior operator is proficient at manipulating the controls or supervising the manipulation of. controls prior to performing licensed duties." A regulatory position (#24) of this proposed revision to Regulatory Guide 1.33 which discusses administrative procedures has. included guidance consistent with the proposed rule change. Work on the proposed rule change and the proposed revision to Regulatory Guide 1.33 concerning these administrative procedures will be coordinated to maintain consistency in the final guidance.
9.
An NRC staff effort is currently underway to evaluate the quality assurance program requirements for the operation phase of nuclear power plants regarding the independence of the quality assurance organization from operating pressures.
For example, the effectiveness of an organizational structure where l
the onsite quality assurance organization reports functionally to offsite management versus reporting functionally to the Plant Manager is being evaluated by the NRC staff.
In-regard.to these: requirements, Appendix 3, " Quality Assurance Criteria l
for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Fart 50,
'" Domestic Licensing of Production and Utilization Facilities,"5 states in Criterion I, Organization:
"The persons and organizations performing quality I'
- 00cument -for' sale by'the Superintendent of Documents, U.S. Government Printing Office, Washington, D.C. 20402.
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.. c, assurance functions-shall have sufficient authority and organizational freedom to identify quality problems; to initiate, recommend, or provide solutions; and to verify implementations of solutions.
Such persons and organizations
. performing quality assurance functions shall report to a management level such that this required authority and organizational freedom, including sufficient independence from cost'and scnedule when opposed to safety considerations, are provided.
Because of the many variables involved, such as the number of person-nel, the type of activity #eing performed, and the location or locations where activities are performed, the organizational structure for axecuting the quality assurance program may take various forms provided that the persons and organiza-tions assigned the quality assurance functions have this required authori,ty and organizational freedom.
Irrespective of the organizational structure, the individual (s) assigned the responsibility for assuring effective execution of any portion of the quality assurance program at any location where activities subject to this' Appendix are,being performed shall have direct access to such levels of management as may be necessary to perform this function."
An NRC document entitled " Guidance on Quality Assurance Requirements During the Operations Phase of Nuclear Power Plants"4 dated October 26, 1973 indicates several organizational structures that are acceptable'to the NRC staff for providing independence of the quality assurance organization from operating pressures.
Furthermore, Section 3.3, " Authorities and Responsibilities for Adminis-trative Controls and Quality Assurance Program Activities," of the draft ANS l
3.2 standard states:
"The persons or organizations responsible for defining and measuring the overall effectiveness of the program shall be designated, shall be sufficiently independent from cost and' scheduling considerations when opposed to safety considerations, shall have direct access to responsible managment at a level
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where action appropriate to the mitigation of safety related quality l
assurance concerns can be accomplished, and shall report regularly on the effectiveness of the program to the plant manager and'the cognizant offsite management.
Persons or organizatiens perf:rming functions of assuring that the administrative controls and quality assurance program is established and implemented or of assuring that an activity has been correctly to:
identify l
quality problems; initiate, recommend or provide solutions, through designated 8
channels; and verify implementation of solutions." In the distribution letter that accompanies this proposed revision to the regulatory guide, public comment is specifically requested on the quality assurance program requirements in this area. The results of this request will be used in the NRC staff determination of the need for revision of the requirements for independence of the quality assurance organization from operating pressures.
There has been some uncertainty with regard to the NRC staff's position when a regulatory guide endorses, as an acceptable method, the " guidelines" a~s well as the " requirements" included in a standard. Where conformance to the recommendations of this regulatory guide is indicated in an application without further qualification, this indicates the applicant will comply with the
" requirements of the draft ANS 3.2 standard, as supplemented or modified by the regulatory position of this guide.
C.
REGULATORY POSITION The overall quality assurance program requirements for the operational phase that are included in Draft 5 (dated August 1980) of the revision to ANSI N18.7/ANS-3.2 are acceptable to the NRC staff and provide an adequate basis for complying with the quality assurance program requiren.3nts of Appendix B to 10 CFR Part 50, subject to the following:
- 1. a.
Throughout the draft revision to ANSI N18.7/ANS-3.2, other documents required to be included as a part of this standard are identified at the point of reference. The specific acceptability of these standards listed in the draft revision to ANSI N18.7/ANS-3.2 has been addressed in the latest revision of the following regulatory guides:
ANSI Standard Regulatory Guide N45.2 1.28 N45.2.1-1.37
-N45.2.2 1.38 N45.2.3 1.39 N45.2.4 1.30 N45.2.5 1.94 N45.2.6 1.58 N45.2.8 1.116 N45.2.9 1.88 N45.2.10 1.74 N45.2.11 1.64 9
N45.2.12 1.144 N45.2.13 1.122 N18.1
- 1. 8 N18.17 1.17 N101.4 1.54 NQA-1 under development b.
Section 3.4.2 states that personnel whose qualifications do not meet those specified in N18.1 and who are performing inspection, examination, and testing activities during the operations phase of the plant, including preoper-ational and startup testing, shall be qualified to in accordance with the requirements of American National Standard Quality Assurance Program Requirements for Nuclear Power Plants NQA-1. However, Regulatcry Guides 1.8, " Personnel.
Selection and Training," and 1.58, "Qualificatibn of Nuclear Power Plant Inspection, Examination, and Testing Personnel," contain guidance for the qualification of these personnel.
c.
Section 5.3.6, " Radiation Control Procedures," contains guidance con-cerning the preparation of procedures to provide for the implementation of a radiation control program. Regulatory Guide 4.15, " Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment," provides guidance on the development of a method acceptable to the NRC staff for designing a program to. ensure the quality of the results of measurements of radioactive materials in the effluents and the environment of nuclear facilities during normal operations.
d.
Section 5.3.9.3, " Procedures for Implementing Emergency Plan," lists elements that are required to be included in the implementing procedures for emergency plan actions.
Additional guidance for the preparation of emergency plan implementing procedures is available in NUREG-0654, " Criteria for Prepar-ation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," dated January 1980.
2.
Section 3.3, " Authorities and Responsibilities for Administrative Controls and Quality Assurance Program Activities," states that persons or organizations responsible for defining and measuring the overall effectiveness of the quality assurance program shall be sufficiently independent from cost and scheduling considerations when opposed to safety considerations.
In addi-tion, those persons or organizations responsible for the areas of training and radiation protection should be independent from operating pressures.
Section 3.3 also states that in those situations where the review funtions are performed 10 4
by personnel not from the quality assurance organization, the quality assurance organization shall review and concur in the procedure associated with the review activities and sha11 perform audits to assure that the review activities have been properly accomp.ished.
In addition, the quality assurance organization should review and concur in the selection of personnel who perform the review.
3.
Section 3.4.2, " Requirements for the Onsite Operating Organization,"
discusses the qualification of the plant staff and the requirements of the operating organization in this area.
In addition to the various fields in which individuals in the onsite operating organization are required to be knowledgeable by Section 3.4.2, the onsite operating organization should include individuals knowledgeable in heat transfer, fluid flow, and thermodynamics.
4.
Section 3.4.3, " Technical Support for the On-Duty Operating Staff,"
states that technical support personnel shall be available (at the station or on call and capable of responding to the plant within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) for the purpose of providing technical advice to the shift supervisor on a 24-hour-a-day basis.
In lieu of the 2-hour time interval allowed in the draft standard, technical support personnel should be capable of arriving at the plant within the time interval specified in NUREG 0654.
5.
Section 4, " Reviews and Audits," establishes requirements for reviews and audits of activities affecting plant safety.
In addition to the organiza-tions addressed in this section, an Independent Safety Engineering Group should be established. Appendix A should be used to establish the principal functions of the Independent Safety Engineering Group and whether the activity should be performed onsite. In addition to the guidance of Appendix A, the ISEG should review the assessments required in Section 5.2.7 of the draft ANS 3.2 standard.
-Regarding the review of procedures by the ISEG, these reviews should address the adequacy of procedures with respect to location of controls and instrumen-tation in the control room, modification of procedures as a result of control room review and changes, and plant operating experience.
Section 4.3.4, " Subjects Requiring Independent Review," lists subjects that are required to be reviewed by the independent review body.
Item (4) requires that violations, deviations and reportable events which require report-ing to the NRC in writing be reviewed by the independent review body.
In addi-tion, violations and deviations identified in NRC inspection reports should be reviewed by the ISEG. Also, plant recor.ds of.aall equipment repairs, adjustments, pev o.
y and replacement.s should be reviewed-en : cont 4nuous4atts to determine trends in the performance of personnel, systems, and components.
From these reviews, j
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the ISEG should evaluate the need for design changes, replacement of components, training improvements, and procedure revision. Upon identification of such a need, the ISEG should interface.with appropriate management and the cognizant plant organization to ensure satisfactory correction of the problem.
Further-more, the ISEG should review the disposition of nonconforming items.
6.
Section 4.3.2.3, " Quorum," discusses the. requirements for a quorum for formal meetings of the standing committee responsible for the independent review program.
In addition, each committee should formally convene a quorum to act on tasks.in their areas of responsibility.
7.
Section 4.5, " Audit Program," states that a comprehensive system of planned and documented audits shall be carried out to verify compliance with all aspects of the administrative controls and quality assurance program. The quality assurance organization should assure that audits are performed in accordance with quality assurance program requirements.
Section 4.5 also lists specific elements that must be audited at increased frequencies.
In Item (3) of the list, the audit of performance of the facility staff should include training records and supervisory evaluations.
8.
Section 5.2.1.4, " Transfer of Responsibility," discusses rules of practice to be established for the transfer of responsibility from one shift to another. Section 5.2.1.4 also provides guidance to assure that personnel or succeeding shifts have a clear understanding of the responsibilities they are assuming and the condition of equipment and systems for which they will assume responsibility.
In addition, procedures should be prepared that require a detailed explana-
' tion of plant status.to the oncoming shift crew.
The encoming shift supervisor should perform a short tour of the plant prior to assuming the duty station with special attention paid to ongoing maintenance and surveillance testing.
The oncoming shift should report to their duty station at least one-half hour prior to shift change to complete and sign a relief turnover checklist. The following items, as a minimum,-should be included in the checklist:
(1) Assurance that specified plant parameters are within allowable limits
-(parameters and allowable limits should be listed on the checklist),
(2) Assurance of the availability and proper alignment of all systems essential to the prevention and mitigation of operational transients and accidents by.- a check of the duty station control console (s) (what to check and criteria for acceptable status should be included on the checklist),
12
'(3) Identification ~ of systems 'and components that are in'a degraded mode-of operation permitted by the Technical Specifications. For such systems and components, the length of time in the degraded mode should be compared with-the Technical. Specifications action statement and should be recorded as a. separate entry.on the enecklist.
9.
Section 5.2.1.6', " Human Factors Considerations," establishes actual work time limitations for certain plant personnel.
In' addition, these limita-tions should be applied to all operating personnel who maintain or operate any structures, systems, or components important to safety.
In lieu of item (2) of Section 5.2.1,6, there should be at least a 12-hour break between all work X'
periods.
These intervals ~should be considered maximum working time limits and not a normal working schedule.
Each operating organization should plan to have adequate personnel such that it should not routinely schedule overtime,to com-pensate for inadequate numbers of personnel. Approval of the Station Manager should be obtained on an individual basis when operating personnel will be on-duty at their particular duty stations for longer than.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in a 24-hour period.
i
- 10. Section 5.2.2, " Procedure. Adherence," states that one of the two individuals required to approve changes to procedures which may affect the operational status of plant systems or equipment must be a member of plant supervision holding a senior operator license on the unit affected.
- However,
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changes to procedures should be approved by two members of the plant management staff, at least one of whom holds a senior reactor operator license on the unit affected.
11.
Section 5.2.6, "Equioment Control," establishes requirements for the proper control of plant systems or equipment released for maintenance.
Section 5.2.6 states that release of plant systems or equipment. requires the permission of designated operating personnel holding a senior operator license.
Permission to release plant.sys'tems or equipment'for maintenance or surveillance
-tests should only be granted by the on-duty shift supervisor.
Section 5.-2.6 states that procedures shall require control measures such as locking or tagging to. secure and identify. equipment in a controlled status. Verification by a Tsecond qualified person of correct implementation of. measures provided for
. control of equipment should'be performed.in all instances with the allo'wable
-exception.of. cases where.high radiation exposure may result and, in that case,.
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other means may be used.
Control room tagouts should be designed and installed to prevent obstruction of other. instruments, controls or indicating lights.
In aedition to the requirements regarding the return to ser'vice of equipment important to safety that are discussed in Section 5.2.6, permission to return plant systems or equipment to service should only be granted by the on-duty shift supervisor. Furthermore, proper alignment should be verified by a second qualified person unless all equipment, valves, and switches involved in the activity can be proven to be in their correct alignment by functional testing without adversely affecting the safety of the plant.
The qualified person who performs the verification of correct implementa-tion of equipment control measures or proper al,ignment prior to returning l
equipment to service should be qualified to perform such tasks for the parti-I cular systems involved, should possess operating knowledge of the particular l
systems involved and their relationship to plant safety and should hold a valid reactor operator or senior reactor operator license.
12.
Section 5.2.8, " Surveillance Testing and Inspection Program," states that surveillance tests and inspections shall be conducted and/or witnessed by l
suitably qualified personnel.
In addition, where the surveillance testing and inspections are performed by personnel who are not members of the quality assurance organization, the i
quality assurance organizatii1 should review and concur in the selection of personnel who perform the surveillance testing and inspections, should review and concur in the procedures asscciated with the surveillance testing and 1
inspections, and should perform audits to assure that the surveillance testing and inspections have been accomplished in accordance with those procedures.
13.
Section 5.2.13.1, " Procurement Oc:ument Control," states that pro-curement documents shall require suppliers to provide a quality assurance l
program consistent with th'e pertinent requirements of American National Standard, Quality Assurance Program Requirements for Nuclear Power Plants NQA-1 to the l
extent necessary.
In addition, procurement documents should specify the extent that, suppliers should comply with ANSI N45.2 and the applicable N45.2 series
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standards.
- 14. Section 5.2.15, " Review, Approval and Control of Procedures," states l
that' measures provided_to control and coordinate the approval and issuance of
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documents which prescribe activities affecting quality shall assure that these
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documents, including revisions or changes, are reviewed for adequacy by appro-priately qualified personnel and approved for release by authorized personnel.
In addition to the above requirements, the quality assurance organization should assure that these documents have been prepared, reviewed, and approved in accor-dance with established procedures.
Final approval provisions of these documents should include review and approval by two members of the plant management staff, at least one of whom holds a senior reactor operator license on the unit affected.
As part of the initial procedure review, a step-by-step walk through process of all new procedures should be performed by personnel who are to use the parti-cular procedure to ensure that the procedure can be safely implemented.
This section also states that the requirement for routine followup review of plant procedures can be accomplished in several ways, including (but not necessarily limited to): documented step-by-step use of the procedure (such as occurs when the procedure has a step-by-step checkoff associated with it), or detailed scrutiny of the procedure as part of a documented training program, drill, simulator exercise, or other such activity. This section also states that a revision of a procedure constitutes a procedure review.
However, the only method that should be considered acceptable to meet the requirement for routine followup review of plant procedures is the review of procedures by a designated review group as an independent activity that is at least as rigorous as the initial procedure review.
15.
Section 5.2.17, " Inspections," requires that inspection of activities affecting safety be performed by qualified individuals other than those wno performed or directly supervised the activity being inspected. Where tme inspections are performed by personnel who are not members of the quality assur-acce organization, the quality assurance organization should review and concur in the selection of personnel who perform the inspections, should review and concur in the procedure associated with the inspection, and should perform audits to assure that the inspections have been accomplished in accordance with those procedures.
16.
Section 5.2.19.1, "Preoperational Tests," states that checklists shall be used to verify and document that all plant systems are arranged in their correct valve lineup.
In' addition, checklists should be used to verify and document that all. plant systems are arranged in their correct equipment, and switch lineups.
-15
- 17. Section 5.2.19.2, " Tests Prior to and During Initial Plant Operation,"
establishes the responsibilities.and scheduling of the initial startup test program following fuel loading.. n addition to the guidance of Section 5.2.19.2, provisions should be made to ensure equipment, valve, and switch lineups are accomplished and verified following testing and prior to conducting plant startup for operation.
18.
Section 5.3.2, " Procedure Content," states that plant procedures shall include a list of specified elements, as appropriate.
Item (5), Precautions, of Section 5.3.2 states that precautions may be specified separate from the procedures.
If precautions are specified separately, then the procedure should require that the precautions be reviewed prior to commencement of the activity.
19.
Section 5.3.3, " System Procedures," states that system procedures shall contain checkoff lists, where appropriate, which are prepared in suffi-cient detail to assure an adequate verification of the status af the system.
System procedures should contain checklists,'or reference documents that contain checklists, in all instances.
20.
Section 5.3.5, " Maintenance Procedures," describes items that main-tenance procedures are required to contain.
Item (3), Post Maintenance Check l
Out and Return to Service, states that instructions shall be included, or referenced, for. returning the equipment to its normal operating status. A checklist should be included in the maintenance procedure or another document I
that is referenced by the maintenance procedure.
21.
Section 5.3.9, " Emergency Procedures," contains guidance concerning the preparation of procedures to guide operations during potential emergencies.
Any immediate operator actions that are necassary to micigate the consequences of an abnormal plant condition should be memorized by operating personnel regardless of the classification of that condition (emergency, off-normal, etc.).
Section 5.3.9 states that engineered safety features shall not be overridden without careful consideration by the shift supervisor. Also, flow rates associ-ated with engineered safety features should not be changed without careful consideration. by the shift supervisor.
In addition, departure from the emergency procedure should require the prior approval of the senior reactor operator directing operations in the control room.
22.
Section 5.3.9.1, " Emergency Procedure Format and Content," lists elements that are required to be included in emergency procedures.
Item (2),
Symptoms, states that symptoms shall be included to aid in the identification 16 1
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I I
of the emergency..Among-other symptoms, the procedure should specify plant parameters that 'are not expected to change and identification characteristics of degraded core cooling where this is a possibility.
In addition to the examples l'
of immediate actions in Item (4), the operator should confirm that auxiliary systems and the ultimate heat sink'are available and/or operating.
l.
l 23.
Section 5.3.10, " Test and Inspection Procedures," states that require-ments for'cestoration of the system to normal conditions, if applicable, shall be included in test and inspection procedures.
Instructions for the restora-tion of the systen to the condition consistent with the plant operating status should be' included, or. referenced, in test and inspection procedures in all instances.
- 24. 'Section 1, " Administrative Procedures," of Appendix A lists activities that should be covered by written administrative procedures.
In addition, the following activities should be covered by written administrative procedures:
a.
Control of feedback of operating experience to provide plant personnel with timely and useful information.
b.
Classes of Emergency Action Levels:
Notification of Unusual Event, i
Alert, Site Emergency, and General Emergency.
c.
Overtime work by operating personnel who maintain or operate any l-structures, systems, or components important to safety.
d.
Assurance that an operator.or senior operator is proficient at manipulating the controls or supervising the manipulation of controls prior to performing licensed duties.
25.
Section 4, " Procedures for Abnormal, Offnormal, or Alarm Conditions,"
states that, where applicable, such procedures may be written so as to cover more than a single annunciator.
However, each annunciator associated with structures, systems, or components important to safety should have a separate
. written procedure.
- 26..The guidelines (indicated by the verb "should") of the draft revision to ANSI'N18.7/ANS-3.2 contained:in the following sections have suffi-i l~
cient safety importance to be treated the same as the requirements (indicated by the verb "shall") of. the' standard:
a.
.Section 5.1--The guideline concerning coverage-in written procedures l
[
or. orders to' describe the quality assurance program of the typical procedures.
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(
listed in Appendix A.
17
...g.-
b.
Section 5.2.1.6--The guideline concerning authorization of deviations from working hour limitations by the station manager.
c.
Section 5.2.2.--The guideline concerning commitment to memory of immediate actions in emergancy procedures.
d.
Section 5.2.13.3--The guideline (permission) that allows identifica-tion on an item or on records traceable to that item.
e.
Section 5.2.19.1--The guidelines concerning (1) the similarity of preoperational test procedures to those discussed in 5.3.3 and 5.3.4, and (2) the modification of procedures to require variation in control parameters, such as pump stops and restarts, cycling valves and varying flows so that system performance can be evaluated.
f.
Section 5.3--The guideline concerning coverage by written procedures of typical activities addressed in Appendix A.
g.
Section 5.3.9.1--The guideline that describes the automatic actions inSection5.3.9.1(3)N X
h.
Appendix A--The guidelines in the introductory paragraph (page 121) concerning (1) coverage by written procedures of the typical activities addressed in the Appendix, and (2) coverage by procedures of many other activities carried out during the operational phase of nuclear power plants which are not included in the Appendix.
- i. Appendix A, Section 6--The guidelines concerning (1) preparation prior to' beginning work of procedures for the repair or replacement of equipment in Section 6c;g C(2) development of procedures that could be categorized either as maintenance or operating procedures in Section 6d; (3) preparation, before reactor operation is begun, of general administrative procedures for the control of maintenance, repair, replacement, and_ modification work in Section Ge; and (4) areas to be addressed by procedures in Section 6e.
- j. Appendix A, Section 8--The guideline concerning providing proce-dures that delineate on a step-by-step basis the correct calibration method for each type of equipment to be calibrated in Section 8a.
k.'
Appendix A, Section 9--The guideline concerning addressing hand =
ling and storage of chemical reagents, particularly those subject to degrada-tion due to time, temperature, or light, in procedures.
18 t
1 D. IMPLEMENTATION l
The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this regulatory guide.
l This proposed revision to the regulatory guide-has been released to encourage public participation in its development.
Except in those cases in which an applicant proposes an acceptable alternative method for complying wth specified portions of the Commission's regulations, the method to be_ des-cribed in the active guide reflecting public comments will be used in the
~
evaluation of compliance with the Commission's regulations with regard to overall quality assurance program requirements for the operation phase of nuclear power l
plants, beginning on the implementaton date to be specified in the active guide, for operating license applicants. This implementation date for operating license app'licants will in no case be earlier than January 1,1981.
A letter will be sent to each operating plant requesting information on the modifications necessary to'its operating organization in order to comply
(
with this proposed revision to Regulatory Guide 1.33 and the draft ANS 3.2 l
standard. The results of this request will be used in determination of the need for application of this Revision to the Regulatory guide to operating plants and required implementation dates.
It is the intent of the NRC, at this l_
point-in time, to require implementation for opera;ing plants of this revision I
to Regulatory Guide 1.33 within 6 months.after the effective issuance date of this revision to the guide, unless adequate justification for additional time for implementation is provided in response to the above request.
i l
i 19 i
J
APPENDIX A EXCERPT FROM COMMISSION INFORMATION PAPER SECY-80-242 INCLUDING TABLE 1 "The enclosed Table 1 provides a listing of the principal functions now assigned to the plant staff review group (the PORC) and to the utility's independent review and audit group in accordance with current staff guidelines (Standard Review Plan, Regulatory Guide 1.33, Standard Technical Specifications).
In the following discussion, we elaborate the role and characteristics we envision for the new Independent Safety Engineering Group (ISEG). As reflected in the Action Plan, these are our tentative, or working, criteria that are being revised and refined with experience in their application to new OLs prior to their general application to all licensees.
We do not envision that the new Independent Safety Engineering Group would replace either existing review group.
Rather, it would be an additional independent group of five dedicated, full-time engineers, located onsite, but reporting offsite to a high level corporate. official who is not in the management chain for power production.
The Independent Safety Engineering Group will increase the available technical expertise located onsite and will allow for continuing, systematic and independent assessment of plant activities.
Integration of the Shift Technical Advisors into the Independent Safety Engineering Group could enhance the group's contact with and knowledge of day-to-day plant operations and provide additional expertise.
The functions of the Independent Safety Engineering Group require daily contact with the operating personnel and continued access to plant' facilities and records. The independent safety review functions can, therefore, best be carried out by a group p:ysically located onsite. However, for utilities with multiple sites, it may.be possible to perform portions of the independent safety assessment function ~in a centralized location for all of the utility's plants.
In such cases, an onsite group still will be required, but it may be slightly smaller than would be the case _if it were performing the entire 20
l i
independent safety assessment function.
The last column of Table 1 indicates which of the: audit activities must be performed onsite, which could be performed offsite, and which could be performed either onsite or offsite depending upon the utility organizational structure or geographical. configuration.
The function of the ISEG is to examine plant operating characteristics,
. NRC issuances, Licensing Information Service advisories and other appropriate sourr:es which may indicate areas for improving plant safety. Where useful improvements can be achieved, it is expected that this group would develop and l
present deta11ed recommendations for revised procedures, equipment modifications l
or other means. Anqther principal function of the ISEG would be to maintain l
surveillance of plant operations and maintenance activities to provide independ-ent verification that these activities are performed correctly and that human errors are reduced as far as practical.
This is not to suggest detailed l
auditing of operations by the ISEG.
Rather, it is intended that through 1
oversight and utility understanding of safety related operations, ISEG will be in a position to advise utility management on the overall quality of operations.
Thus, ISEG would not be responsible for sign-off functions such that it becomes involved in the operating organization.
l The information provided by industry-wide groups, such as INPO and NSAC, will be valuable'in identifying generic areas of concern.
Information developed by these groups will be made available to the industry. However, determination l
of the impact on each facility will remain a plant specific task that should be conducted or reviewed by the utility staff.
l.
Therefore, in our view, the existence of industry-wide groups does not l
alleviate the need for the ISEG or the other utility review groups."
21
^ ~
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).
I 1
!i l
Table 1
.. Responsibility Current Proposed i
Plant Staff Independent Incependent Review Group Review And Safety (PORC)
Audit Group Engineering l
Group Review procedures and changes to procedures X
On-Site l
Review proposed tests and experiments X
On-Site
-Review changes and modifications to Unit systems and equipment X
On-Site L
Review Safety Evaluations for changes to prncedures, equipment, or systems, or tests and experiments to verify l
that actions do not constitute unre-viewed safety issues X
On-Site Review changes, tests, or experiments
which involve unreviewed safety issues X
On-Site Review changes to technical speciff-cations X
X On-Site Review violations of technical spec-ifications, including reports and recommendations to avoid recurrence X
On-Site Review violations of codes, regula-tio'ns, orders, technical specifica-tions, ifcense requirements, and of internal procedures or instructions X
On-Site l
l
. Review events requiring 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> reporting to NRC X
X On-Site Review unit operations to detect-l potential safety hazards X
On-Site Review significant operating abnor-
-malities, or deviations from normal or expected performance of plant equipment.
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X-On-Site
. 22 a
e-
,-+
w w
l l
l Table 1 (contd:)
i Responsibility Current Proposed i
Plant Staff Independent Independent Review Group Review And Safety (PORC)
Audit Group Engineering Group l
Review unanticipated deficiencies in j
some aspect of design or operation X
Either Review Security Plan and implementing j
proccdures and recommend changes X
Either Review Emergency Plan and implement-ing procedures and recommend changes X
Either Review reports and meeting minutes of other review groups and provide oversight X
Off-Site Assure corrective action and recom-mendations of other review groups are implemented X
Off-Site Other reviews as requested by the Offsite Review Group X
On-Site Assess Plant Staff Performance On-Site Evaluate effectiveness of QA Program Either Evaluate operating experience of the unit and units of similar i
design On-Site l
Other reviews deemed necessary by an independent reviewer X
Either r
23
,,