Regulatory Guide 1.33: Difference between revisions

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{{Adams
{{Adams
| number = ML13109A458
| number = ML003739995
| issue date = 06/13/2013
| issue date = 02/28/1978
| title = Rev 3, Quality Assurance Program Requirements (Operation).
| title = Quality Assurance Program Requirements (Operation)
| author name =  
| author name =  
| author affiliation = NRC/RES
| author affiliation = NRC/RES
Line 9: Line 9:
| docket =  
| docket =  
| license number =  
| license number =  
| contact person = Rodriguez-Luccioni H L
| contact person =  
| case reference number = DG-1300
| document report number = RG-1.33, Rev 2
| document report number = RG-1.033, Rev 3
| package number = ML13109A437
| document type = Regulatory Guide
| document type = Regulatory Guide
| page count = 6
| page count = 8
}}
}}
{{#Wiki_filter:Written suggestions regarding this guide or development of new guides may be submitted through the NRC's public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html.    Electronic copies of this regulatory guide, previous versions of this guide, and other recently issued guides are available through the NRC's public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/.  The regulatory guide is also available through the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession No. ML13109A458.  The regulatory analysis may be found in ADAMS under Accession No. ML13109A459 and the staff responses to the public comments on DG-1300 may be found under ADAMS Accession No. ML13109A467.
{{#Wiki_filter:U.S. NUCLEAR REGULATORY
COMMISSION
Revlklon 2 February 1978 REGULATORY
GUIDE OFFICE OF STANDARDS
DEVELOPMENT
REGULATORY
GUIDE 1.33 QUALITY ASSURANCE
PROGRAM REQUIREMENTS (OPERATION)


U.S. NUCLEAR REGULATORY COMMISSION June 2013 Revision 3 REGULATORY GUIDE OFFICE OF NUCLEAR REGULATORY RESEARCH
==A. INTRODUCTION==
REGULATORY GUIDE 1.33 (Draft was issued as DG-1300, dated January 2013)  
Appendix B, "Quality Assurance Criteria for Nu clear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50, "Licensing of Production and Utili zation Facilities," establishes quality assurance re quirements for the operation of nuclear power plant safety-related structures, systems, and components.
QUALITY ASSURANCE PROGRAM REQUIREMENTS (OPERATION)
 
This regulatory guide describes a method acceptable to the NRC staff for complying with the Commis sion's regulations with regard to overall quality as surance program requirements for the operation phase of nuclear power plants. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.
 
==B. DISCUSSION==
Subcommittee ANS-3,' Reactor Operations, of the American Nuclear Society Standards Committee de veloped ANSI N18.7-1972, which contained criteria for administrative controls for nuclear power plants dur ing operation.
 
This standard, along with ANSI N45.2-1971, "Quality Assurance Program Require ments 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 require ments of Appendix B to 10 CFR Part 50; 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 define the general quality assurance program "requirements" for the operation phase. This revised standard was approved by the American National Standards Com mittee NI8, Nuclear Design Criteria.
 
It was sub sequently approved and designated N18.7-1976/
-*Lines indicate substantive changes from previous issue.ANS-3.2, "Administrative Controls and Quality As surance for the Operational Phase of Nuclear Power Plants," by the American National Standards Insti tute on February 19, 1976.  There had been some uncertainty with regard to the NRC staff's position when a regulatory guide en dorses, as an acceptable method, the "guidelines" as well as the "requirements" included in a standard.
 
The NRC staff has evaluated the guidelines contained in N18.7-1976/ANS-3.2 with respect to importance to safety. Revision I of this regulatory guide clarified the NRC staff's position on the "requirements" and "guidelines" included in ANSI N18.7-1976/ANS
3.2. Where conformance to the recommendations of this regulatory guide is indicated in an application without further qualification, this indicates the appli cant will comply with the "requirements" of ANSI N18.7-1976/ANS-3.2, as supplemented or modified by the regulatory position of this guide.  Section 1, "Scope," of ANSI NI8.7-1976/ANS
3.2 states that this standard contains criteria for ad ministrative controls and quality assurance for nu clear power plants during the operational phase of plant life and that this phase is generally considered to commence with initial fuel loading, except for cer tain preoperational activities.
 
In this regard, a sepa rate regulatory guide addressing the quality assurance program for the preoperational phase will be issued.  Other regulatory guides may be issued or this regula tory guide may be revised, if necessary, to amplify the general requirements contained in this standard.
 
Appendix A to this guide has been further revised as a result of additional'
comments received on the guide and additional staff review.  ,Copies may be obtained from American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 6052
 
===5. USNRC REGULATORY ===
GUIDES Comments shoukl be sent to the Secretary of the Commission.
 
US. Nuclear Regu Regulatory Guides ore issued to describe and make available to the Public methods latory Commission, Washington, D.C. 20555, Attention:
Docketing and Service acceptable to the NRC staff of implementing specific parts of the Commission's regulations, to delineate techniques used by the staff in evaluating specific problems The guides are issued in the following ten broad divisions or postulated accidents, or to Provide guidance to applicants.
 
Regulatory Guides awt not substitutes for regulations, and compliance with them is not required.
 
1. Power Reaclors 6. TProducts Methods and solutions different from those set out in the guides will be accept. 2. Research and Test Reactors 7. Trancsortation aWle if they provide a basis for the findings requisite to the issuance or continuance
3. Fuels and Materials Facilities a. Occupational Health o f a p ecr mit or license aby th e Co mni, son. 4. Environm etnt l and Siting 9. Antitrust Review 5. Materials and Plant Protection t0. General Comments end suggestions for improvements in these guides we encouraged at all Requests for single copies of issued guides iwhich may be reiroduced)  
or for place tires, and guides will be revised, as appropriate, to accomnmodate comments and ment on an automatic distribution list for single copies of future guides in specific "to reflect new information or experience.


==A. INTRODUCTION==
This guide was revised as a result of divisions should be made in writing to the US. Nuclear Regulatory Commission.
Purpose  This regulatory guide (RG) describes methods that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for managerial and administrative Quality Assurance (QA) controls to be used for nuclear power plants during operations.


Applicable Rules and Regulations This guide describes methods that the NRC staff considers acceptable for complying with the provisions of regulations in 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," of the Code of Federal Regulations (Ref. 1), §50.34(b)(6)(ii), Contents of applications; technical information  and  10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants," (10 CFR Part 52) (Ref. 2) §52.79(a)(27), Contents of applications; technical information in final safety analysis report. Both sections require compliance with 10 CFR Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," which, in part, requires the establishment of QA controls for the implementation of managerial and administrative controls to assure safe operation.
eubltantive comments received from the Public and additional staff review. Washington, D.C. 20555. Attention Director.


Related Guidance Guidance for the establishment and execution of QA programs for nuclear power plants during their design and construction is in RG 1.28, "Quality Assurance Program Requirements (Design and Construction)," (Ref. 3).  
Division of Document Coistrol.


Rev. 3 of RG 1.33, Page 2 Purpose of Regulatory Guides The NRC issues regulatory guides to describe to the public methods that the staff considers acceptable for use in implementing specific parts of the agency's regulations, to explain techniques that the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to applicants. Regulatory guides are not substitutes for regulations and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis fo r the findings required for the issuance or continuance of a permit or license by the Commission.
C. REGULATORY
POSITION The overall quality assurance program require ments for the operation phase that are included in ANSI N18.7-1976/ANS-3.2 are acceptable to the NRC staff and provide an adequate basis for comply ing with the quality assurance program requirements of Appendix B to 10 CFR Part 50, subject to the fol lowing: I. ANSI N18.7-1976/ANS-3.2 requires the prep aration of many procedures to carry out an effective quality assurance program. Appendix A, "Typical Procedures for Pressurized Water Reactors and Boil ing Water Reactors," to this regulatory guide should be used as guidance to ensure minimum procedural coverage for plant operating activities, including re lated maintenance activities.


Paperwork Reduction Act This regulatory guide contains information collection requirements covered by 10 CFR Part 50 and 10 CFR Part 52 that the Office of Management and Budget (OMB) approved under OMB control numbers 3150-0011 and 3150-0151, respectively.  The NRC may neither conduct nor sponsor, and a person is not required to respond to, an informa tion collection request or requirement unless the requesting document displays a currently valid OMB control number.
Appendix A lists typical safety-related activities that should be covered by written procedures but does not provide a complete listing of needed procedures.


==B. DISCUSSION==
Many other activities carried out during the operation phase of a nuclear power plant require written procedures not included in Appendix A. Appendix A may also contain proce dures that are not applicable to an applicant because of the configuration of the nuclear power plant. The procedures listed in Appendix A may be combined, separated, or deleted to conform to the applicant's procedures plan.  2. Throughout ANSI NI8.7-1976/ANS-3.2, other documents required to be included as a part of this standard are identified at the point of reference.
Reason for Revision This revision (Revision 3) of RG 1.33 endorses ANSI/ANS 3.2-2012, "Managerial, Administrative, and Quality Assurance Controls for Operational Phase of Nuclear Power Plants,"  (Ref. 4).  Revision 2 of RG 1.33 endorsed a previous version of the standard, which was ANS 3.2/ANSI N18.7-1976, "Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants," (Ref. 5).  The updated standard incorporates operational experience since the original standard was developed, and is better focused on QA of plant operations because information on QA of design and construction was moved to another standard.


Background Revision 2 of RG 1.33 (Ref. 6) endorsed ANS 3.2/ANSI 18.7-1976, but required numerous clarifications or modifications of the standard in the RG's Regulatory Position section.  Since then, licensees have obtained NRC approval to use various alternate positions to Revision 2 of RG 1.33.
The specific acceptability of these standards listed in ANSI N18.7-1976/ANS-3.2 has been addressed in the latest revision of the following regulatory guides: ANSI Standard N45.2 N45.2.1 N45.2.2 N45.2.3 N45.2.4 N45.2.5 N45.2.6 N45.2.8 N45.2.9 N45.2. 10 N45.2. I I N45.2.13 N18. 1 N 18.17 N 101.4 Regulatory Guide 1.28 1.37 1.38 1.39 1.30 1.94 1.58 1.116 1.88 1.74 1.64 1.123 1.8 1.17 1.54 Note: N45.2.12 is discussed in NRC documents WASH-1283, "Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants," (Grey Book) and WASH 1309, "Guidance on Quality Assurance Re quirements During the Construction Phase of Nuclear Power Plants," (Green Book)and will be endorsed by a regulatory guide upon its approval as an ANSI standard.


In addition, the American Society of Mechanical Engineers (ASME) issued NQA-1, "Quality Assurance Program Requirements for Nuclear Power Plants," (Ref. 7), which was focused on design and construction issues.  The NRC has endorsed NQA-1 in 10 CFR 50.55a, "Codes and Standards".  The NRC revised a related RG (RG 1.28) to endorse NQA-1-2008 and the NQA-1a-2009 Addenda, "Quality Assurance Requirements for Nuclear Facility Applications."
3. Section 4.3.4, "Subjects Requiring Independ ent Review," Item (3) states, in part, that changes to the technical specifications or license amendments re lated to nuclear safety are required to be reviewed by the independent review body prior to implementation.


ANSI/ANS 3.2-2012 revised ANS 3.2/ANSI 18.7-1976 to remove information related to design and construction to be consistent with NQA-1, and to incorporate the alternate positions approved by the NRC since ANS 3.2/ANSI 18.7-1976 was issued.  Revision 3 of RG 1.33 clarifies the distinction of the quality assurance program during design and construction from those managerial and administrative controls implemented during the operational phase of nuclear power plants.
It should be noted that proposed changes to technical specifications or license amendments should be re viewed by the independent review body prior to their submittal to the Commission for approval.


Rev. 3 of RG 1.33, Page 3 Harmonization with International Standards The International Atomic Energy Agency (IAEA) has established a series of safety guides and standards constituting a high level of safety for protecting people and the environment. IAEA safety guides present international good practices and increasingly reflect best practices to help users striving to achieve high levels of safety.  Pertinent to this regulatory guide, the IAEA Safety Standards, and their Safety Requirement GS-R-3, "The Management System for Facilities and Activities," (Ref. 8), issued in 2006, address administrative and quality assurance controls for the operational phase of nuclear power plants.  This regulatory guide incorporates similar administrative and quality assurance controls for the operational phase and is consistent with the basic safety principles provided in the IAEA Safety Standard.
4. Section 4.5, "Audit Program," of ANSI N18.7-1976/ANS_3.2 states that audits of selected aspects of operational phase activities shall be per formed with a frequency commensurate with their safety significance and in such a manner as to ensure that an audit of all safety-related functions is com pleted within a period of 2 years. In amplification of this requirement, the following program elements should be audited at the indicated frequencies:
a. The results of actions taken to correct de ficiencies that affect nuclear safety and occur in facil ity equipment, structures, systems, or method of operation-at least once per 6 months.  b. The conformance of facility operation to pro visions contained within the technical specifications and applicable license conditions-at least once per 12 months.  c. The performance, training, and qualifications of the facility staff-at least once per 12 months.  5. The guidelines (indicated by the verb "should")
of ANSI N18.7-1976/ANS-3.2 contained in the following sections have sufficient safety impor tance to be treated the same as the requirements (in dicated by the verb "shall") of the standard:
a. Section 4.4-The guidelines concerning re view activities of the onsite operating organization, except the guideline that refers to screening subjects of potential concern.


Documents Discussed in Staff Regulatory Guidance This regulatory guide endorses the use of one or more voluntary consensus codes or standards developed by external organizations.  These codes or standards may contain references to other codes or standards. These references should be considered individually. If a referenced standard has been incorporated separately into NRC regulations, licensees and applicants must comply with that standard as set forth in the regulation.  If the referenced standard has been endorsed in a regulatory guide, the standard constitutes a method acceptable to the NRC staff for meeting a regulatory requirement as described in the specific regulatory guide. If a refere nced standard has been neither incorporated into NRC regulations nor endorsed in a regulatory guide, licensees and applicants may consider and use the information in the referenced standard, if appropriate ly justified and consistent with current regulatory practice.
b. Section 5.2.3-The guideline concerning re view and updating of standing orders.  c. Section 5.2.4-The guideline concerning re view, updating, and cancellation of special orders.  d. Section 5.2.7. 1-The guidelines that address adequate design and testing of replacement parts.  e. Section 5.2.13.4-The guideline concerning special handling tools and equipment.


C. STAFF REGULATORY GUIDANCE  The requirements included in ANSI/ANS 3.2-2012, "Managerial, Administrative and Quality Assurance Controls for the Operational Phase of Nuclear Power Plants", for implementation during the operation phase of nuclear power plants, are acceptable to the NRC staff and provide an adequate basis for complying with the requirements of Appendix B to 10 CFR Part 50, subject to the following condition on the use of ANSI/ANS 3.2-2012:
f. Section 5.2.19(2)-The guideline for check ing plant operating procedures during the testing pro gram.  g. Section 5.2.19. 1-The guidelines for preop erational tests, except the guideline that refers to a 1.33-2 run-in period for equipment.


* ANSI/ANS 3.2-2012 requires the preparation of many procedures to carry out an effective QA program.  Appendix A of ANSI/ANS 3.2-2012, "Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors," should be used as guidance to assure the minimal procedural coverage for plant operating activities, including related maintenance activities.  Appendix A lists typical safety-related activities that should be covered by written procedures, but does not provide a complete listing of necessary procedures. Many other activities carried out during the operation phase of a nuclear power plant require written procedures, which may or may not be applicable, because of the configuration of the nuclear power plant.  The procedures listed in Appendix A may be added to, combined, separated or deleted to conform to the applicant's procedure plan.
In addition to these guidelines, the prerequisite steps for each equipment
-test should be completed prior to the commencement of the preoperational test.  h. Section 5.3.2-The guidelines that describe the content (excluding format) of procedures, except for the guidelines that address (1) a separate state ment of applicability in Section 5.3.2(2), (2) inclu sion of references in procedures, as applicable, in Section 5.3.2(3), and (3) inclusion of quantitative control guides in Section 5.3.2(6).   
i.. Section 5.3.9-The guideline concerning emergency procedures requiring prompt implementa tion of immediate operator actions when required to prevent or mitigate the consequences of a serious condition.


Rev. 3 of RG 1.33, Page 4  
j. Section 5.3.9.1-The guidelines that describe the content (excluding format) for.' the title in Section 5.3.9.1 (1); the inclusion of symptoms to aid in iden-tification in Section 5.3.9.1(2);
automatic actions in Section 5.3.9.1(3);
immediate operator action, excluding those guidelines contained in the examples, in Section 5.3.9.1(4);
and subsequent operator ac tions in Section 5.3.9.1(5). 


==D. IMPLEMENTATION==
==D. IMPLEMENTATION==
The purpose of this section is to provide information on how applicants and licensees
The purpose of this section is to provide informa tion to applicants and licensees regarding the NRC staff's plans for using this regulatory guide.  This guide reflects current NRC practice.
1 may use this guide and information regarding the NRC's plans for using this regulatory guide.  In addition, it describes how the NRC staff complies with 10 CFR 50.109, "Backfitting" and any applicable finality provisions in 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants."  Use by Applicants and Licensees Applicants and licensees may voluntarily
 
2 use the guidance in this document to demonstrate compliance with the underlying NRC regulations. Methods or solutions that differ from those described in this regulatory guide may be deemed acceptable if they provide sufficient basis and information for the NRC staff to verify that the proposed alternative demonstrates compliance with the appropriate NRC regulations. Current licensees may continue to use Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2 for complying with the identified regulations as long as their current licensing basis remains unchanged.   Licensees may use the information in this regulatory guide for actions which do not require NRC review and approval such as changes to a facility design under 10 CFR 50.59, "Changes, Tests, and Experiments."  Licensees may use the information in this regulatory guide or applicable parts to resolve regulatory or inspection issues. Use by NRC Staff  The NRC staff does not intend or approve any imposition or backfitting of the guidance in this regulatory guide.  The NRC staff does not expect any existing licensee to use or commit to using the guidance in this regulatory guide, unless the licensee makes a change to its licensing basis. The NRC staff does not expect or plan to request licensees to voluntarily adopt this regulatory guide to resolve a generic regulatory issue. The NRC staff does not expect or plan to initiate NRC regulatory action which would require the use of this regulatory guide. Examples of such unplanned NRC regulatory actions include issuance of an order requiring the use of the regulatory guide, requests for information under 10 CFR 50.54(f) as to whether a licensee intends to commit to use of this regulatory guide, generic communication, or promulgation of a rule requiring the use of this regulatory guide without further backfit consideration. During regulatory discussions on plant specific operational issues, the staff may discuss with licensees various actions consistent with staff positions in this regulatory guide, as one acceptable means of meeting the underlying NRC regulatory requirement. Such discussions would not ordinarily be considered backfitting even if prior versions of this regulatory guide are part of the licensing basis of the facility. However, unless this regulatory guide is part of the licensing basis for a facility, the staff may not represent to the licensee that the licensee's failure to comply with the positions in this regulatory guide constitutes a violatio
There fore, except in those cases in which the applicant proposes an acceptable alternative method for com plying with the specified portions of the Commis sion's regulations, the method described herein is being and will continue to be used in the evaluation of submittals for operating license applications until this guide is revised as a result of suggestions from the public or additional staff review.1.33-3 APPENDIX A TYPICAL PROCEDURES
FOR PRESSURIZED
WATER REACTORS AND BOILING WATER REACTORS The following are typical safety-related activities that should be covered by written procedures.
 
This appendix is not intended as an inclusive listing of all needed procedures since many other activities carried out during the operation phase of nuclear.power plants should be covered by procedures not included in this list.  1. Administrative Procedures a. Security and Visitor Control b. Authorities and Responsibilities for Safe Opera tion and Shutdown c. Equipment Control (e.g., locking and tagging) d. Procedure Adherence and Temporary Change Method e. Procedure Review and Approval f. Schedule for Surveillance Tests and Calibration g. Shift and Relief Turnover h. Log Entries, Record Retention, and Review Procedures i. Access to Containment j. Bypass of Safety Functions and Jumper Control k. Maintenance of Minimum Shift Complement and Call-In of Personnel I. Plant Fire Protection Program m. Communication System Procedures
2. General Plant Operating Procedures a. Cold Shutdown to Hot Standby b. Hot Standby to Minimum Load (nuclear start up) c. Recovery from Reactor Trip d. Operation at Hot Standby e. Turbine Startup and Synchronization of Generator f. Changing Load and Load Follow (if applicable)
g. Power Operation and Process Monitoring h. Power Operation with less than Full Reactor Coolant Flow i. Plant Shutdown to Hot Standby j. Hot Standby to Cold Shutdown k. Preparation for Refueling and Refueling Equipment Operation I. Refueling and Core Alterations
3. Procedures for Startup, Operation, and Shutdown of Safety-Related PWR Systems Instructions for energizing, filling, venting, drain ing, startup, shutdown, and changing modes of oper ation should be prepared, as appropriate, for the fol lowing systems: a. Reactor Coolant System b. Control Rod Drive System (including part length rods) c. Shutdown Cooling System d. Emergency Core Cooling System e. Component Cooling Water System
 
====f. Containment ====
(1) Maintaining Containment Integrity
(2) Special Containment Systems (a) Atmosphere (b) Subatmospheric (c) Double-Wall Containment with Controlled Interspace (d) Ice Condenser
(3) Containment Ventilation System (4) Containment Cooling System g. Atmosphere Cleanup Systems h. Fuel Storage Pool Purification and Cooling Sys tem i. Main Steam System j. Pressurizer Pressure and Spray Control Systems k. Feedwater System (feedwater pumps to steam generator)
1. Auxiliary Feedwater System m. Service Water System n. Chemical and Volume Control System (includ ing Letdown/Purification System) o. Auxiliary or Reactor Building Heating and Ventilation p. Control Room Heating and Ventilation q. Radwaste Building Heating and Ventilation r. Instrument Air System s. Electrical System (1) Offsite (access circuits)
(2) Onsite (a) Emergency Power Sources (e.g., diesel generator, batteries) (b) A.C. System (c) D.C. System t. Nuclear Instrument System (1) Source Range (2) Intermediate Range (3) Power Range (4) Incore System u. Reactor Control and Protection System
 
====v. Hydrogen Recombiner ====
4. Procedure for Startup, Operation, and Shutdown of Safety-Related BWR Systems 1.33-4 Instructions for energizing, filling, venting, drain ing, startup, shutdown, and changing modes of oper ation should be prepared, as appropriate, for the following systems: a. Nuclear Steam Supply System (Vessel and Recirculating System) b. Control Rod Drive System c. Reactor Cleanup System d. Liquid Poison System (Standby Liquid Con trol System) e. Shutdown Cooling and Reactor Vessel Head Spray System f. High Pressure Coolant Injection g. Reactor Core Isolation Cooling System h. Emergency Core Cooling Systems i. Closed Cooling Water System
 
====j. Containment ====
(1) Maintaining Integrity
(2) Containment Ventilation System (3) Inerting and deinerting k. Fuel Storage Pool Purification and Cooling System I. Main Steam System (reactor vessel to turbine) m. Turbine-Generator System n. Condensate System (hotwell to feedwater pumps, including demineralizers and resin regeneration)
o. Feedwater System (feedwater pumps to reactor vessel) p. Makeup System (filtration, purification, and water transfer)
q. Service Water System r. Reactor Building Heating and Ventilation Systems s. Control Room Heating and Ventilation Systems t. Radwaste Building Heating and Ventilation Systems u. Standby Gas Treatment System v. Instrument Air System w. Electrical System (1) Offsite (access circuits)
(2) Onsite (a) Emergency Power Sources (e.g., diesel generator, batteries) (b) A.C. System (c) D.C. System x. Nuclear Instrument System (1) Source Range (2) Intermediate Range (3) Power Range (4) TIP System y. Reactor Protection System z. Rod Worth Minimizer 5. Procedures for Abnormal, Offnormal, or Alarm Conditions Since these procedures are numerous and corre spond to the number of alarm annunciators, the pro cedures are not individually listed. Each safety related annunciator should have its own written procedure, which should normally contain (1) the meaning of the annunciator, (2) the source of the sig nal, (3) the immedate action that is to occur automat ically, (4) the immediate operation action, and (5) the long-range actions.
 
6. Procedures for Combating Emergencies and Other Significant Events a. Loss of Coolant (including significant PWR steam generator leaks) (inside and outside primary containment) (large and small, including leak-rate determination)
b. Loss of Instrument Air c. Loss of Electrical Power (and/or degraded power sources) d. Loss of Core Coolant Flow e. Loss of Condenser Vacuum f. Loss of Containment Integrity g. Loss of Service Water h. Loss of Shutdown Cooling i. Loss of Component Cooling System and Cool ing to Individual Components j. Loss of Feedwater or Feedwater System Failure k. Loss of Protective System Channel 1. Mispositioned Control Rod or Rods (and rod drops) m. Inability to Drive Control Rods n. Conditions Requiring Use of Emergency Bora tion or Standby Liquid Control System o. Fuel Cladding Failure or High Activity in Reactor Coolant or Offgas p. Fire in Control Room or Forced Evacuation of Control Room q. Turbine and Generator Trips r. Other Expected Transients that may be Applicable s. Malfunction of Automatic Reactivity Control System t. Malfunction of Pressure Cdntrol System u. Reactor Trip v. Plant Fires w. Acts of Nature (e.g., tornado, flood, dam failure, earthquakes)
x. Irradiated Fuel Damage While Refueling y. Abnormal Releases of Radioactivity z. Intrusion of Demineralizer Resin Into Primary System (BWR Plants) 7. Procedures for Control of Radioactivity (For limit ing materials released to environment and limiting personnel exposure)1.33-5 I
a. Liquid Radioactive Waste System ( I) Collection.
 
Demineralizing.
 
Filtering, Evaporating and Concentrating, and Neutralizing
(2) Sampling and Monitoring
(3) Discharging to Effluents b. Solid Waste System (1) Spent Resins and Filter Sludge Handling (2) Baling Machine Operation
(3) Drum Handling and Storage c. PWR Gaseous Effluent System (1) (2) (3) (4)Collection.
 
Storage. and Discharge Sampling and Monitoring Air Ejector and Stack Monitoring Ventilation Air Monitoring d. BWR Air Extraction.
 
Offgas Treatment.
 
and Other Gaseous Effluent Systems (1) (2) (3) (4) (5)Mechanical Vacuum Pump Operation Air Ejector Operation Packing Steam Exhauster Operation Sampling Air Ejector. Ventilation.
 
and Stack Monitor e. Radiation Protection Procedures (I) Access Control to Radiation Areas Includ.  ing a Radiation Work Permit System (2) Radiation Sur\e\s (3) Airborne Radioacti
 
====v. ity Monitoring ====
(4) Contamination Control (5) Respiratory Protection
(6) Training in Radiation Protection
(7) Personnel Monitoring
(8) Bioassay Program (9) Implementation of ALARA Program f. Area Radiation Monitoring S)stem Operation g. Process Radiation Monitoring System Operation h. Meteorological Monitoring
8. Procedures for Control of Measuring and Test Equipment and for Surveillance Tests, Proce dures. and Calibrations a. Procedures of a type appropriate to the cir cumstances should be provided to ensure that tools, gauges. instruments, controls, and other measuring and testing devices are properly controlled, calib rated. and adjusted at specified periods to maintain accuracy.
 
Specific examples of such equipment to be calibrated and tested are readout instruments, inter lock permissive and prohibit circuits, alarm devices, sensors. signal conditioners, controls, protective cir cuits, and laboratory equipment.
 
b. Specific procedures for surveillance tests, in spections.
 
and calibrations should be written (imple menting procedures are required for each surveillance test. inspection, or calibration listed in the technical specifications):
(1) Pressurized Water Reactors (a) Containment Leak-Rate Tests (b) Containment Isolation Tests (c) Containment Local Leak Detection Tests (d) Containment Heat and Radioactivity Removal Systems Tests (e) Containment Tendon Tests and Inspections (f) Service Water System Functional Tests (g) Main Steam Isolation Valve Tests (hi Fire Protection System Functional Tests (i) Boric Acid Tanks-Level Instrumenta tion Calibrations (j) Emergency Core Cooling System Tests (k Control Rod Operability and Scram Time Tests (1) Reactor Protection System Tests and Calibrations (m) Permissi',es-Tests and Calibrations (n) Refueling System Circuit Tests (o) Emergency Boration System Functional Tests (p) DNB Checks and Incore-Excore Flux Monitor Correlations (q) Emergency Power Tests tr) Auxiliary Feedwater System Tests (s) NSSS Pressurization and Leak Detection (t) Inspection of Reactor Coolant System Pressure Boundary (u) Inspection of Pipe Hanger Settings (v) Control Rod Drive System Functional Tests (w) Heat Balance-Flux Monitor Calibrations Ix) Pressurizer and Main Steam Safety Valve. Tests (y) Leak Deection Systems Tests (z) Axial and Radial Flux Pattern Determi nations (aa) Area. Portable.


====n.     ====
and Airborne Radiation M on it or Calibrations (bb) Process Radiation Monitor Calibrations (cc) Environmental Monitor Calibrations (dd) Safety Valve Tests (ee) Turbine Overspeed Trip Tests (ff) Water Storage Tanks-Level In strumentation Calibration
1 In this section, "licensees" refers to licensees of nuclear power plants under 10 CFR Parts 50 and 52; and the term "applicants," refers to applicants for licenses and permits for (or relating to) nuclear power plants under 10 CFR Parts 50 and 52, and applicants for standard design approvals and standard design certifications under 10 CFR Part 52. 2  In this section, "voluntary" and "voluntarily" means that the licensee is seeking the action of its own accord, without the force of a legally binding requirement or an NRC representation of further licensing or enforcement action.
(2) Boiling Water Reactors (a) Containment Leak-Rate and Penetration Leak-Rate Tests (b) Containment Isolation Tests (c) Containment Vacuum Relief Valve Tests (d) Containment Spray System Tests (e) Standby Gas Treatment System Tests (including filter tests)1.33-6 (f) Service Water System Functional Tests (g) Main Steam Isolation Valve Tests (h) Fire Protection System Functional Tests (i) Nitrogen Inerting System Tests (j) Emergency Core Cooling System Tests (k) Control Rod Operability and Scram Time Tests (1) Reactor Protection System Tests and Calibrations (m) Rod Blocks-Tests and Calibrations (n) Refueling System Circuit Tests (o) Liquid Poison System Tests (p) Minimum Critical Heat Flux Checks and Incore Flux Monitor Calibrations (q) Emergency Power Tests (r) Isolation Condenser or RCIC Tests (s) NSSS Pressurization and Leak Detection (t) Inspection of Reactor Coolant System Pressure Boundary (u) Inspection of Pipe Hanger Settings (v) Control Rod Drive System Functional Tests (w) Heat Balance (x) Autoblowdown System Tests (y) Leak Detection System Tests (z) Axial and Radial Flux Pattern Determi nations (aa) Area, Portable, and Airborne Radiation Monitor Calibrations (bb) Process Radiation Monitor Calibrations I (cc) Environmental Monitor Calibrations (dd) Safety Valve Tests (ee) Turbine Overspeed Trip Test (ff) Water Storage Tanks-Level In strumentation Calibrations (gg) Reactor Building Inleakage Tests 9. Procedures for Performing Maintenance a. Maintenance that can affect the performance of safety-related equipment should be properly pre planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.


Rev. 3 of RG 1.33, Page 5 If an existing licensee voluntarily seeks a license amendment or change and (1) the NRC staff's consideration of the request involves a regulatory issue directly relevant to this new or revised regulatory guide and (2) the specific subject matter of this regulatory guide is an essential consideration in the staff's determination of the acceptability of the licensee's request, then the staff may request that the licensee either follow the guidance in this regulatory guide or provide an eq uivalent alternative process that demonstrates compliance with the underlying NRC regulatory requirements. This is not considered backfitting as defined in 10 CFR 50.109(a)(1) or a violation of any of the issue finality provisions in 10 CFR Part 52.  Additionally, an existing applicant may be required to adhere to new rules, orders, or guidance if 10 CFR 50.109(a)(3) applies.  If a licensee believes that the NRC is either using this regulatory guide or requesting or requiring the licensee to implement the methods or processes in this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfit appeal with the NRC in accordance with the guidance in NUREG-1409, "Backfitting Guidelines" (Ref. 9) and NRC Management Directive 8.4, "Management of Facility-specific Backfitting and Information Collection" (Ref. 10).     
Skills normally possessed by qualified maintenance personnel may not require detailed step-by-step delineation in a pro cedure. The following types of activities are among those that may not require detailed step-by-step writ ten procedures:
Rev. 3 of RG 1.33, Page 6 REFERENCES
(1) Gasket Replacement
3  1. U.S. Code of Federal Regulations, 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities." 
(2) Trouble-Shooting Electrical Circuits (3) Changing Chart or Drive Speed Gears or Slide Wires on Recorders b. Preventive maintenance schedules should be developed to specify lubrication schedules, inspec tions of equipment, replacement of such items as fil ters and strainers, and inspection or replacement of parts that have a specific lifetime such as wear rings.c. Procedures for the repair or replacement of equipment should be prepared prior to beginning work. Such procedures for major equipment that is expected to be repaired or replaced during the life of the plant should preferably be written early in plant life. The following are examples of such procedures for major equipment:
2. U.S. Code of Federal Regulations, 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants." 
(1) Repair of PWR Steam Generator Tubes (2) Replacement and Repair of Control Rod Drives (3) (4) Filters (5) (6) (7)Replacement of Recirculation Pump Seals Replacement of Important Strainers and Repair or Replacement of Safety Valves Repair of Incore Flux Monitoring System Replacement of Neutron Detectors d. Procedures that could be categorized either as maintenance or operating procedures should be developed for the following activities.


3. Regulatory Guide 1.28, "Quality Assurance Program Criteria (Design and Construction), "U.S. Nuclear Regulatory Commission, Washington, DC.
Instructions for these activities may be included in systems procedures.


4. American National Standards Institute (ANSI)/ American Nuclear Society (ANS) 3.2-2012, "Managerial, Administrative, and Quality Assurance Controls for Operational Phase of Nuclear Power Plants."
(1) Exercise of equipment that is normally idle but that must operate when required (2) Draining and Refilling Heat Exchangers
5. American Nuclear Society
(3) Draining and Refilling Recirculation Loop (4) Draining and Refilling the Reactor Vessel (5) Draining and Refilling Steam Generators
4 (ANS) 3.2/ American National Standards Institute
(6) Removal of Reactor Head (7) Disconnection and Reconnection of Wiring Penetrating Reactor Vessel Head (8) Demineralizer Resin Regeneration or Replacement e. General procedures for the control of mainte nance, repair, replacement, and modification work should be prepared before reactor operation is begun. These procedures should include information on areas such as the following:
5 (ANSI) 18.7-1976, "Administrative Controls and Quality A
(1) Method for obtaining permission and clear ance for operation personnel to work and for logging such work and (2) Factors to be taken into account, including the necessity for minimizing radiation exposure to workmen, in preparing the detailed work procedures.
ssurance for the Operational Phase of Nuclear Power Plants."
6. Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," Revision 2, U.S. Nuclear Regulatory Commission, Washington, DC.


7. American National Standards Institute (ANSI)/American Society of Mechanical Engineers (ASME) NQA-1, "Quality Assurance Program Requirements for Nuclear Power Plants."
10. Chemical and Radiochemical Control Procedures Chemical and radiochemical control procedures should be written to prescribe the nature and fre quency of sampling and analyses, the instructions maintaining water quality within prescribed limits, and the limitations on concentrations of agents that may cause corrosive attack or fouling of heat-transfer surfaces or that may become sources of radiation hazards due to activation.
6  8. International Atomic Energy Agency Safety Requirement GS-R-3, "The Management Systems for Facilities and Activities," issued 2006.


7  9. NUREG 1409, "Backfitting Guidelines," U.S. Nuclear Regulatory Commission, Washington, DC.  10. Management Directive 8.4, "Management of Facility-specific Backfitting and information Collection," U.S. Nuclear Regulatory Commission, Washington, DC.
These procedures should specify laboratory instructions and calibration of lab oratory equipment.


3 Publicly available NRC published documents are available electronically through the NRC Library on the NRC's public Web site at: http://www.nrc.gov/reading-rm/doc-collections/.  The documents can also be viewed on-line or printed for a fee in the NRC's Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, Washington, DC 20555; telephone 301-415-4737 or (800) 397-4209; fax (301) 415-3548; and e-mail pdr.resource@nrc.gov
Extreme importance must be placed on laboratory procedures used to determine 1.33-7 concentration and species of radioactivity in liquids and gases prior to release, including representative UNITED STATES NUCLEAR REGULATORY
4 Copies of American Nuclear Society (ANS) standards may be purchased from the ANS Web site (http://www.new.ans.org/store/); or by writing to: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60526, U.S.A., Telephone 800-323-3044.
COMMISSION
WASHINGTON, D.C. 20555-0001 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 sampling, validity of calibration techniques, and ade. quacy of analyses.


5 Copies of American National Standard s Institute (ANSI) standards may be purchased from ANSI, 1819 L Street, NW., Washington, DC 20036, on their Web site at http://webstore.ansi.org/; telephone (202) 293-8020; fax (202) 293-9287; or e-mail storemanager@ansi.org.    6 Copies of American Society of Mechanical Engineers (ASME) standards may be purchased from ASME, Two Park Avenue, New York, New York 10016-5990; telephone (800) 843-2763.  Purchase information is available through the ASME Web-based store at http://www.asme.org/Codes/Publications/
FIRST CLASS MAIL POSTAGE AND FEES PAID USNRC PERMIT NO. G-67 PRNTED ON RECYCLED PAPER}}
. 7 Copies of International Atomic Energy Agency(IAEA) standards may be purchased from IAEA, Wagramerstrasse 5, P.O. Box 100, A-1400 Vienna, Austria; Telephone: +43 1 2600 22529 (or 22530). Purchase information is available through the ASME
Web-based store at http://www-pub.iaea.org/MTCD/publications/publications.asp
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Revision as of 03:41, 21 September 2018

Quality Assurance Program Requirements (Operation)
ML003739995
Person / Time
Issue date: 02/28/1978
From:
Office of Nuclear Regulatory Research
To:
References
RG-1.33, Rev 2
Download: ML003739995 (8)
v • d • e


U.S. NUCLEAR REGULATORY

COMMISSION

Revlklon 2 February 1978 REGULATORY

GUIDE OFFICE OF STANDARDS

DEVELOPMENT

REGULATORY

GUIDE 1.33 QUALITY ASSURANCE

PROGRAM REQUIREMENTS (OPERATION)

A. INTRODUCTION

Appendix B, "Quality Assurance Criteria for Nu clear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50, "Licensing of Production and Utili zation Facilities," establishes quality assurance re quirements for the operation of nuclear power plant safety-related structures, systems, and components.

This regulatory guide describes a method acceptable to the NRC staff for complying with the Commis sion's regulations with regard to overall quality as surance program requirements for the operation phase of nuclear power plants. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position.

B. DISCUSSION

Subcommittee ANS-3,' Reactor Operations, of the American Nuclear Society Standards Committee de veloped ANSI N18.7-1972, which contained criteria for administrative controls for nuclear power plants dur ing operation.

This standard, along with ANSI N45.2-1971, "Quality Assurance Program Require ments 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 require ments of Appendix B to 10 CFR Part 50; 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 define the general quality assurance program "requirements" for the operation phase. This revised standard was approved by the American National Standards Com mittee NI8, Nuclear Design Criteria.

It was sub sequently approved and designated N18.7-1976/

-*Lines indicate substantive changes from previous issue.ANS-3.2, "Administrative Controls and Quality As surance for the Operational Phase of Nuclear Power Plants," by the American National Standards Insti tute on February 19, 1976. There had been some uncertainty with regard to the NRC staff's position when a regulatory guide en dorses, as an acceptable method, the "guidelines" as well as the "requirements" included in a standard.

The NRC staff has evaluated the guidelines contained in N18.7-1976/ANS-3.2 with respect to importance to safety. Revision I of this regulatory guide clarified the NRC staff's position on the "requirements" and "guidelines" included in ANSI N18.7-1976/ANS

3.2. Where conformance to the recommendations of this regulatory guide is indicated in an application without further qualification, this indicates the appli cant will comply with the "requirements" of ANSI N18.7-1976/ANS-3.2, as supplemented or modified by the regulatory position of this guide. Section 1, "Scope," of ANSI NI8.7-1976/ANS

3.2 states that this standard contains criteria for ad ministrative controls and quality assurance for nu clear power plants during the operational phase of plant life and that this phase is generally considered to commence with initial fuel loading, except for cer tain preoperational activities.

In this regard, a sepa rate regulatory guide addressing the quality assurance program for the preoperational phase will be issued. Other regulatory guides may be issued or this regula tory guide may be revised, if necessary, to amplify the general requirements contained in this standard.

Appendix A to this guide has been further revised as a result of additional'

comments received on the guide and additional staff review. ,Copies may be obtained from American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 6052

5. USNRC REGULATORY

GUIDES Comments shoukl be sent to the Secretary of the Commission.

US. Nuclear Regu Regulatory Guides ore issued to describe and make available to the Public methods latory Commission, Washington, D.C. 20555, Attention:

Docketing and Service acceptable to the NRC staff of implementing specific parts of the Commission's regulations, to delineate techniques used by the staff in evaluating specific problems The guides are issued in the following ten broad divisions or postulated accidents, or to Provide guidance to applicants.

Regulatory Guides awt not substitutes for regulations, and compliance with them is not required.

1. Power Reaclors 6. TProducts Methods and solutions different from those set out in the guides will be accept. 2. Research and Test Reactors 7. Trancsortation aWle if they provide a basis for the findings requisite to the issuance or continuance

3. Fuels and Materials Facilities a. Occupational Health o f a p ecr mit or license aby th e Co mni, son. 4. Environm etnt l and Siting 9. Antitrust Review 5. Materials and Plant Protection t0. General Comments end suggestions for improvements in these guides we encouraged at all Requests for single copies of issued guides iwhich may be reiroduced)

or for place tires, and guides will be revised, as appropriate, to accomnmodate comments and ment on an automatic distribution list for single copies of future guides in specific "to reflect new information or experience.

This guide was revised as a result of divisions should be made in writing to the US. Nuclear Regulatory Commission.

eubltantive comments received from the Public and additional staff review. Washington, D.C. 20555. Attention Director.

Division of Document Coistrol.

C. REGULATORY

POSITION The overall quality assurance program require ments for the operation phase that are included in ANSI N18.7-1976/ANS-3.2 are acceptable to the NRC staff and provide an adequate basis for comply ing with the quality assurance program requirements of Appendix B to 10 CFR Part 50, subject to the fol lowing: I. ANSI N18.7-1976/ANS-3.2 requires the prep aration of many procedures to carry out an effective quality assurance program. Appendix A, "Typical Procedures for Pressurized Water Reactors and Boil ing Water Reactors," to this regulatory guide should be used as guidance to ensure minimum procedural coverage for plant operating activities, including re lated maintenance activities.

Appendix A lists typical safety-related activities that should be covered by written procedures but does not provide a complete listing of needed procedures.

Many other activities carried out during the operation phase of a nuclear power plant require written procedures not included in Appendix A. Appendix A may also contain proce dures that are not applicable to an applicant because of the configuration of the nuclear power plant. The procedures listed in Appendix A may be combined, separated, or deleted to conform to the applicant's procedures plan. 2. Throughout ANSI NI8.7-1976/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 ANSI N18.7-1976/ANS-3.2 has been addressed in the latest revision of the following regulatory guides: ANSI Standard N45.2 N45.2.1 N45.2.2 N45.2.3 N45.2.4 N45.2.5 N45.2.6 N45.2.8 N45.2.9 N45.2. 10 N45.2. I I N45.2.13 N18. 1 N 18.17 N 101.4 Regulatory Guide 1.28 1.37 1.38 1.39 1.30 1.94 1.58 1.116 1.88 1.74 1.64 1.123 1.8 1.17 1.54 Note: N45.2.12 is discussed in NRC documents WASH-1283, "Guidance on Quality Assurance Requirements During Design and Procurement Phase of Nuclear Power Plants," (Grey Book) and WASH 1309, "Guidance on Quality Assurance Re quirements During the Construction Phase of Nuclear Power Plants," (Green Book)and will be endorsed by a regulatory guide upon its approval as an ANSI standard.

3. Section 4.3.4, "Subjects Requiring Independ ent Review," Item (3) states, in part, that changes to the technical specifications or license amendments re lated to nuclear safety are required to be reviewed by the independent review body prior to implementation.

It should be noted that proposed changes to technical specifications or license amendments should be re viewed by the independent review body prior to their submittal to the Commission for approval.

4. Section 4.5, "Audit Program," of ANSI N18.7-1976/ANS_3.2 states that audits of selected aspects of operational phase activities shall be per formed with a frequency commensurate with their safety significance and in such a manner as to ensure that an audit of all safety-related functions is com pleted within a period of 2 years. In amplification of this requirement, the following program elements should be audited at the indicated frequencies:

a. The results of actions taken to correct de ficiencies that affect nuclear safety and occur in facil ity equipment, structures, systems, or method of operation-at least once per 6 months. b. The conformance of facility operation to pro visions contained within the technical specifications and applicable license conditions-at least once per 12 months. c. The performance, training, and qualifications of the facility staff-at least once per 12 months. 5. The guidelines (indicated by the verb "should")

of ANSI N18.7-1976/ANS-3.2 contained in the following sections have sufficient safety impor tance to be treated the same as the requirements (in dicated by the verb "shall") of the standard:

a. Section 4.4-The guidelines concerning re view activities of the onsite operating organization, except the guideline that refers to screening subjects of potential concern.

b. Section 5.2.3-The guideline concerning re view and updating of standing orders. c. Section 5.2.4-The guideline concerning re view, updating, and cancellation of special orders. d. Section 5.2.7. 1-The guidelines that address adequate design and testing of replacement parts. e. Section 5.2.13.4-The guideline concerning special handling tools and equipment.

f. Section 5.2.19(2)-The guideline for check ing plant operating procedures during the testing pro gram. g. Section 5.2.19. 1-The guidelines for preop erational tests, except the guideline that refers to a 1.33-2 run-in period for equipment.

In addition to these guidelines, the prerequisite steps for each equipment

-test should be completed prior to the commencement of the preoperational test. h. Section 5.3.2-The guidelines that describe the content (excluding format) of procedures, except for the guidelines that address (1) a separate state ment of applicability in Section 5.3.2(2), (2) inclu sion of references in procedures, as applicable, in Section 5.3.2(3), and (3) inclusion of quantitative control guides in Section 5.3.2(6).

i.. Section 5.3.9-The guideline concerning emergency procedures requiring prompt implementa tion of immediate operator actions when required to prevent or mitigate the consequences of a serious condition.

j. Section 5.3.9.1-The guidelines that describe the content (excluding format) for.' the title in Section 5.3.9.1 (1); the inclusion of symptoms to aid in iden-tification in Section 5.3.9.1(2);

automatic actions in Section 5.3.9.1(3);

immediate operator action, excluding those guidelines contained in the examples, in Section 5.3.9.1(4);

and subsequent operator ac tions in Section 5.3.9.1(5).

D. IMPLEMENTATION

The purpose of this section is to provide informa tion to applicants and licensees regarding the NRC staff's plans for using this regulatory guide. This guide reflects current NRC practice.

There fore, except in those cases in which the applicant proposes an acceptable alternative method for com plying with the specified portions of the Commis sion's regulations, the method described herein is being and will continue to be used in the evaluation of submittals for operating license applications until this guide is revised as a result of suggestions from the public or additional staff review.1.33-3 APPENDIX A TYPICAL PROCEDURES

FOR PRESSURIZED

WATER REACTORS AND BOILING WATER REACTORS The following are typical safety-related activities that should be covered by written procedures.

This appendix is not intended as an inclusive listing of all needed procedures since many other activities carried out during the operation phase of nuclear.power plants should be covered by procedures not included in this list. 1. Administrative Procedures a. Security and Visitor Control b. Authorities and Responsibilities for Safe Opera tion and Shutdown c. Equipment Control (e.g., locking and tagging) d. Procedure Adherence and Temporary Change Method e. Procedure Review and Approval f. Schedule for Surveillance Tests and Calibration g. Shift and Relief Turnover h. Log Entries, Record Retention, and Review Procedures i. Access to Containment j. Bypass of Safety Functions and Jumper Control k. Maintenance of Minimum Shift Complement and Call-In of Personnel I. Plant Fire Protection Program m. Communication System Procedures

2. General Plant Operating Procedures a. Cold Shutdown to Hot Standby b. Hot Standby to Minimum Load (nuclear start up) c. Recovery from Reactor Trip d. Operation at Hot Standby e. Turbine Startup and Synchronization of Generator f. Changing Load and Load Follow (if applicable)

g. Power Operation and Process Monitoring h. Power Operation with less than Full Reactor Coolant Flow i. Plant Shutdown to Hot Standby j. Hot Standby to Cold Shutdown k. Preparation for Refueling and Refueling Equipment Operation I. Refueling and Core Alterations

3. Procedures for Startup, Operation, and Shutdown of Safety-Related PWR Systems Instructions for energizing, filling, venting, drain ing, startup, shutdown, and changing modes of oper ation should be prepared, as appropriate, for the fol lowing systems: a. Reactor Coolant System b. Control Rod Drive System (including part length rods) c. Shutdown Cooling System d. Emergency Core Cooling System e. Component Cooling Water System

f. Containment

(1) Maintaining Containment Integrity

(2) Special Containment Systems (a) Atmosphere (b) Subatmospheric (c) Double-Wall Containment with Controlled Interspace (d) Ice Condenser

(3) Containment Ventilation System (4) Containment Cooling System g. Atmosphere Cleanup Systems h. Fuel Storage Pool Purification and Cooling Sys tem i. Main Steam System j. Pressurizer Pressure and Spray Control Systems k. Feedwater System (feedwater pumps to steam generator)

1. Auxiliary Feedwater System m. Service Water System n. Chemical and Volume Control System (includ ing Letdown/Purification System) o. Auxiliary or Reactor Building Heating and Ventilation p. Control Room Heating and Ventilation q. Radwaste Building Heating and Ventilation r. Instrument Air System s. Electrical System (1) Offsite (access circuits)

(2) Onsite (a) Emergency Power Sources (e.g., diesel generator, batteries) (b) A.C. System (c) D.C. System t. Nuclear Instrument System (1) Source Range (2) Intermediate Range (3) Power Range (4) Incore System u. Reactor Control and Protection System

v. Hydrogen Recombiner

4. Procedure for Startup, Operation, and Shutdown of Safety-Related BWR Systems 1.33-4 Instructions for energizing, filling, venting, drain ing, startup, shutdown, and changing modes of oper ation should be prepared, as appropriate, for the following systems: a. Nuclear Steam Supply System (Vessel and Recirculating System) b. Control Rod Drive System c. Reactor Cleanup System d. Liquid Poison System (Standby Liquid Con trol System) e. Shutdown Cooling and Reactor Vessel Head Spray System f. High Pressure Coolant Injection g. Reactor Core Isolation Cooling System h. Emergency Core Cooling Systems i. Closed Cooling Water System

j. Containment

(1) Maintaining Integrity

(2) Containment Ventilation System (3) Inerting and deinerting k. Fuel Storage Pool Purification and Cooling System I. Main Steam System (reactor vessel to turbine) m. Turbine-Generator System n. Condensate System (hotwell to feedwater pumps, including demineralizers and resin regeneration)

o. Feedwater System (feedwater pumps to reactor vessel) p. Makeup System (filtration, purification, and water transfer)

q. Service Water System r. Reactor Building Heating and Ventilation Systems s. Control Room Heating and Ventilation Systems t. Radwaste Building Heating and Ventilation Systems u. Standby Gas Treatment System v. Instrument Air System w. Electrical System (1) Offsite (access circuits)

(2) Onsite (a) Emergency Power Sources (e.g., diesel generator, batteries) (b) A.C. System (c) D.C. System x. Nuclear Instrument System (1) Source Range (2) Intermediate Range (3) Power Range (4) TIP System y. Reactor Protection System z. Rod Worth Minimizer 5. Procedures for Abnormal, Offnormal, or Alarm Conditions Since these procedures are numerous and corre spond to the number of alarm annunciators, the pro cedures are not individually listed. Each safety related annunciator should have its own written procedure, which should normally contain (1) the meaning of the annunciator, (2) the source of the sig nal, (3) the immedate action that is to occur automat ically, (4) the immediate operation action, and (5) the long-range actions.

6. Procedures for Combating Emergencies and Other Significant Events a. Loss of Coolant (including significant PWR steam generator leaks) (inside and outside primary containment) (large and small, including leak-rate determination)

b. Loss of Instrument Air c. Loss of Electrical Power (and/or degraded power sources) d. Loss of Core Coolant Flow e. Loss of Condenser Vacuum f. Loss of Containment Integrity g. Loss of Service Water h. Loss of Shutdown Cooling i. Loss of Component Cooling System and Cool ing to Individual Components j. Loss of Feedwater or Feedwater System Failure k. Loss of Protective System Channel 1. Mispositioned Control Rod or Rods (and rod drops) m. Inability to Drive Control Rods n. Conditions Requiring Use of Emergency Bora tion or Standby Liquid Control System o. Fuel Cladding Failure or High Activity in Reactor Coolant or Offgas p. Fire in Control Room or Forced Evacuation of Control Room q. Turbine and Generator Trips r. Other Expected Transients that may be Applicable s. Malfunction of Automatic Reactivity Control System t. Malfunction of Pressure Cdntrol System u. Reactor Trip v. Plant Fires w. Acts of Nature (e.g., tornado, flood, dam failure, earthquakes)

x. Irradiated Fuel Damage While Refueling y. Abnormal Releases of Radioactivity z. Intrusion of Demineralizer Resin Into Primary System (BWR Plants) 7. Procedures for Control of Radioactivity (For limit ing materials released to environment and limiting personnel exposure)1.33-5 I

a. Liquid Radioactive Waste System ( I) Collection.

Demineralizing.

Filtering, Evaporating and Concentrating, and Neutralizing

(2) Sampling and Monitoring

(3) Discharging to Effluents b. Solid Waste System (1) Spent Resins and Filter Sludge Handling (2) Baling Machine Operation

(3) Drum Handling and Storage c. PWR Gaseous Effluent System (1) (2) (3) (4)Collection.

Storage. and Discharge Sampling and Monitoring Air Ejector and Stack Monitoring Ventilation Air Monitoring d. BWR Air Extraction.

Offgas Treatment.

and Other Gaseous Effluent Systems (1) (2) (3) (4) (5)Mechanical Vacuum Pump Operation Air Ejector Operation Packing Steam Exhauster Operation Sampling Air Ejector. Ventilation.

and Stack Monitor e. Radiation Protection Procedures (I) Access Control to Radiation Areas Includ. ing a Radiation Work Permit System (2) Radiation Sur\e\s (3) Airborne Radioacti

v. ity Monitoring

(4) Contamination Control (5) Respiratory Protection

(6) Training in Radiation Protection

(7) Personnel Monitoring

(8) Bioassay Program (9) Implementation of ALARA Program f. Area Radiation Monitoring S)stem Operation g. Process Radiation Monitoring System Operation h. Meteorological Monitoring

8. Procedures for Control of Measuring and Test Equipment and for Surveillance Tests, Proce dures. and Calibrations a. Procedures of a type appropriate to the cir cumstances should be provided to ensure that tools, gauges. instruments, controls, and other measuring and testing devices are properly controlled, calib rated. and adjusted at specified periods to maintain accuracy.

Specific examples of such equipment to be calibrated and tested are readout instruments, inter lock permissive and prohibit circuits, alarm devices, sensors. signal conditioners, controls, protective cir cuits, and laboratory equipment.

b. Specific procedures for surveillance tests, in spections.

and calibrations should be written (imple menting procedures are required for each surveillance test. inspection, or calibration listed in the technical specifications):

(1) Pressurized Water Reactors (a) Containment Leak-Rate Tests (b) Containment Isolation Tests (c) Containment Local Leak Detection Tests (d) Containment Heat and Radioactivity Removal Systems Tests (e) Containment Tendon Tests and Inspections (f) Service Water System Functional Tests (g) Main Steam Isolation Valve Tests (hi Fire Protection System Functional Tests (i) Boric Acid Tanks-Level Instrumenta tion Calibrations (j) Emergency Core Cooling System Tests (k Control Rod Operability and Scram Time Tests (1) Reactor Protection System Tests and Calibrations (m) Permissi',es-Tests and Calibrations (n) Refueling System Circuit Tests (o) Emergency Boration System Functional Tests (p) DNB Checks and Incore-Excore Flux Monitor Correlations (q) Emergency Power Tests tr) Auxiliary Feedwater System Tests (s) NSSS Pressurization and Leak Detection (t) Inspection of Reactor Coolant System Pressure Boundary (u) Inspection of Pipe Hanger Settings (v) Control Rod Drive System Functional Tests (w) Heat Balance-Flux Monitor Calibrations Ix) Pressurizer and Main Steam Safety Valve. Tests (y) Leak Deection Systems Tests (z) Axial and Radial Flux Pattern Determi nations (aa) Area. Portable.

and Airborne Radiation M on it or Calibrations (bb) Process Radiation Monitor Calibrations (cc) Environmental Monitor Calibrations (dd) Safety Valve Tests (ee) Turbine Overspeed Trip Tests (ff) Water Storage Tanks-Level In strumentation Calibration

(2) Boiling Water Reactors (a) Containment Leak-Rate and Penetration Leak-Rate Tests (b) Containment Isolation Tests (c) Containment Vacuum Relief Valve Tests (d) Containment Spray System Tests (e) Standby Gas Treatment System Tests (including filter tests)1.33-6 (f) Service Water System Functional Tests (g) Main Steam Isolation Valve Tests (h) Fire Protection System Functional Tests (i) Nitrogen Inerting System Tests (j) Emergency Core Cooling System Tests (k) Control Rod Operability and Scram Time Tests (1) Reactor Protection System Tests and Calibrations (m) Rod Blocks-Tests and Calibrations (n) Refueling System Circuit Tests (o) Liquid Poison System Tests (p) Minimum Critical Heat Flux Checks and Incore Flux Monitor Calibrations (q) Emergency Power Tests (r) Isolation Condenser or RCIC Tests (s) NSSS Pressurization and Leak Detection (t) Inspection of Reactor Coolant System Pressure Boundary (u) Inspection of Pipe Hanger Settings (v) Control Rod Drive System Functional Tests (w) Heat Balance (x) Autoblowdown System Tests (y) Leak Detection System Tests (z) Axial and Radial Flux Pattern Determi nations (aa) Area, Portable, and Airborne Radiation Monitor Calibrations (bb) Process Radiation Monitor Calibrations I (cc) Environmental Monitor Calibrations (dd) Safety Valve Tests (ee) Turbine Overspeed Trip Test (ff) Water Storage Tanks-Level In strumentation Calibrations (gg) Reactor Building Inleakage Tests 9. Procedures for Performing Maintenance a. Maintenance that can affect the performance of safety-related equipment should be properly pre planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.

Skills normally possessed by qualified maintenance personnel may not require detailed step-by-step delineation in a pro cedure. The following types of activities are among those that may not require detailed step-by-step writ ten procedures:

(1) Gasket Replacement

(2) Trouble-Shooting Electrical Circuits (3) Changing Chart or Drive Speed Gears or Slide Wires on Recorders b. Preventive maintenance schedules should be developed to specify lubrication schedules, inspec tions of equipment, replacement of such items as fil ters and strainers, and inspection or replacement of parts that have a specific lifetime such as wear rings.c. Procedures for the repair or replacement of equipment should be prepared prior to beginning work. Such procedures for major equipment that is expected to be repaired or replaced during the life of the plant should preferably be written early in plant life. The following are examples of such procedures for major equipment:

(1) Repair of PWR Steam Generator Tubes (2) Replacement and Repair of Control Rod Drives (3) (4) Filters (5) (6) (7)Replacement of Recirculation Pump Seals Replacement of Important Strainers and Repair or Replacement of Safety Valves Repair of Incore Flux Monitoring System Replacement of Neutron Detectors d. Procedures that could be categorized either as maintenance or operating procedures should be developed for the following activities.

Instructions for these activities may be included in systems procedures.

(1) Exercise of equipment that is normally idle but that must operate when required (2) Draining and Refilling Heat Exchangers

(3) Draining and Refilling Recirculation Loop (4) Draining and Refilling the Reactor Vessel (5) Draining and Refilling Steam Generators

(6) Removal of Reactor Head (7) Disconnection and Reconnection of Wiring Penetrating Reactor Vessel Head (8) Demineralizer Resin Regeneration or Replacement e. General procedures for the control of mainte nance, repair, replacement, and modification work should be prepared before reactor operation is begun. These procedures should include information on areas such as the following:

(1) Method for obtaining permission and clear ance for operation personnel to work and for logging such work and (2) Factors to be taken into account, including the necessity for minimizing radiation exposure to workmen, in preparing the detailed work procedures.

10. Chemical and Radiochemical Control Procedures Chemical and radiochemical control procedures should be written to prescribe the nature and fre quency of sampling and analyses, the instructions maintaining water quality within prescribed limits, and the limitations on concentrations of agents that may cause corrosive attack or fouling of heat-transfer surfaces or that may become sources of radiation hazards due to activation.

These procedures should specify laboratory instructions and calibration of lab oratory equipment.

Extreme importance must be placed on laboratory procedures used to determine 1.33-7 concentration and species of radioactivity in liquids and gases prior to release, including representative UNITED STATES NUCLEAR REGULATORY

COMMISSION

WASHINGTON, D.C. 20555-0001 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 sampling, validity of calibration techniques, and ade. quacy of analyses.

FIRST CLASS MAIL POSTAGE AND FEES PAID USNRC PERMIT NO. G-67 PRNTED ON RECYCLED PAPER


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