ML19208B986

Discusses Use of Stds Writing Groups for Implementing long- Term TMI-2 Lessons Learned Task Force Recommendations. Matter to Be Discussed at Earliest Task Force Meeting. Supporting Info Encl

ML19208B986
Person / Time
Site:	Crane
Issue date:	08/15/1979
From:	Mattson R NRC - TMI-2 LESSONS LEARNED TASK FORCE
To:	NRC - TMI-2 LESSONS LEARNED TASK FORCE
References
NUDOCS 7909240261
Download: ML19208B986 (32)

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1. "'%g UNITED STATES

,k NUCLEAR REGULATORY COMMisslON [) e Q WASHINGTON, D. C. 20555 k'3n#?! 3 V AUG 151979 MEMORANDUM FOR: TMI-2 Lessons Learned Task Force FROM: Roger J. Mattson, Director Lessons Learned Task Torce

SUBJECT:

USE OF STANDARDS WRITING GROUPS FOR IMPLEMEf1 TING LONG TERM TMI-2 LESSONS LEARNED TASK FORCE RECOMMENDATIONS On July 19, 1979, representatives of the ANS Nuclear Power Plant Standuds Comittee made a presentation to R. Tedesco, J. Milhoan, and me on some priority standards from their perspective. Attached are handouts provided at the meeting. The first two handouts show tha organization of C.e ANS Standards Comittee and the ANS Nuclear Power Plant Standards Comittee. The remaining handouts describe present activities and omnosed activities of specific ANS standards writing groups. I believe that several of the standards projects described in the attachments are ideally suited as implementation devices for long tenn recommendatior.s we have in mind. This is especially true considering the caliber of technMal talent the ANS can bring to bear through its standards writing groups. I am willing to write a letter to the ANS Nuclear Power Plant Standards Comittee to ask for priority work on any of these topics that the Task Force feels will speed the sort of change we seek to foster. We will discuss the possibility of this course of action at an early Task Force meeting. If we can be specific in what we want, perhaps in a manner analogous to Regula-tory Guide 1.97, I am also willing to initiate letters or memos requesting such work be accomplished on some of these other standards. I do not procose at this time that we devote Task Force attention to developing scope statements for fu-ture projects since we have too much else to do in preparing our final report. The Office of Standards Development will be in a position to prepare such scope statements upon conclusion of our report. m ! \\ -, My Roger J. Mattson, Director Lessons learned Task Force cc: Guy Arlotto, SD 799 022 7909240

AMERICAN NUCLEAR SOCIETY STANDARDS COMMITTEE G. L. Wessman, Chairman J. A. Prestelo. Vice Chairman M. D. Webor G. A. Arlotto It Admin. Secy. V. S. Boyer. N16 N17 N18 N19 N48 NUPPSCO A. D. Callhan W. L. Whittemore G. L. Wessman R. F. Foster R. E. Tomlinson J. F. Mattay Chairman Chairman Chairman Chairman Chairman Chairman I ( ANS-8 ) ( ANS-1 ) ( ANS-2 ) ( ANS-2 ) Groups under ( ANS-6 ) ( ANS-3 } ( ANS-3 ) management of N16 Chairman ( ANS-10 ) ( ANS '] ( ANS-4 )

• SSC Member-at-Large I

( ANS-9] ( ANS-Idj ( ANS-5 )

1. Liaison Member, NRC

( ANS-5 ) ( ANS-11 ) ( ANS-15 ) ( ANS-50 ) ( ANS-51 ) (A -16 ) ( ANS-19 ) ( N18-20 ) ( ANS-52 ) (ANS-65) e ( ANS-53 ) ( ANSs 7] ( ANS-54 ) c2 N ( ANS-18.1] u ( ANS-5d ( ANS-18.2 ) ( ANS-56 ) ( ANS 19 2 ) ( N18 20 ) D L

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/Alle as ..*=r 4 I. ANSI /ANS-31-1978,"AmericanNationalStandardforSelectionand Training of Nuclear Power Plant Personnel" II. This standard provides criteria for the selection and training of personnel for stationary nuclear power plants. It addresses itself ta the qualifications, responsibilities, and training of personnel in openting and support organizations appropriate for the safe and efficient operation of nuclear power plants. III. Attached are guidelines and statements of policy being follcwed in revising this standard. IV. An ANS-3 group review of this standard revised as indicated will be held July 31-August 1, 1979 Contingent upon the outccme of this review, the standard could be introduced into the formal review and approval cycle by August 15, 1979 H. 7. Gr2cn 7/16/79 ~ 799 025 a-3

P f3 D f n ENCLOSURE U uJ d JI r? D q r f t Guidelines for Revision ~of ANSI /ANS 3.1-197 y di_ C _ k a Paragraph numbering and headings refer to specific sections in ANSI /ANS 3.1-1978.

2. - Definitions A definition for "high school dipicca or ivalent" will be added to clarify t

that the General Educational and Development Test (GED) is the only recognized equivslent for a high school education. Nu_ clear power plant experience. The experience under item 1 will be restricted t a maxi =us of one year. Item 4 vill be revised to clarify that the training must be at an operating nuclear reactor. 3.1 - General The second paragraph,in this section will be replaced by the following. d The qualifications of th'e' personnel filling positions due to the abscnce of a principal shall as a mini =us possess the qualific-tions of the naxt lower level in that field. This does not apply to positione requiring senior reactor operator or reactor operator licensces. The individual who replaces the reactor engineer or the radiation protection g:oup leader shall have a B.S. degree and 2 years' experience, one of which will be nuclear power plant experience or an advunced degree and 1 year nuclear power plant experience. Shall have been onsite for 6 conths. The individuals who replace the instrumentation and.ontrol or chemistry and radiochemistry group.' 2aders shall have a sinirus of three years' experience in their ficid, of which one year shall be at an operating nuclear power plant and 6 months at the site. 4.2.1 - Plant "anagers These paragraphs will be rewritten to include the following requircuents. The plant manager or his designated alternate sha.11 hoid a senior reac: )r operator's license. The plant manager shall have a minimum of ten years' e:cperience at an cperating power plant. The designated alternate to the plant manager shall have a mininua of eight years' experience at an operating pcwer plant. 799 026

\\ G3"f1 9 .j u[ o j] lj ~510 llA Either the plant manager or his designated alternate slf e i of four years of supervisory responsibilley at an oper.d h. 9 eW hbeh plant. Both the plant manager and his delegated alternate shall have a B.S. degree or above in an engineering or scientific field generally associated with power plants. ~ 4.2.2, 4.2.3, and 4.2.4 - Operations, Maintenance, end Technical Managers These sections will be revised to require the manager at a new plant to be onsite six months prior to the coc=encement of preoperational testing. Experience require =ents must be met at the ti=e of cot =encement of preop-testing or appoint =ent to position whichever is later. 4.4.4 - Radiation Protection This section will be revised to require the radiation protection engineer to be located onsite. 4.4.5 - Ouality Assurance -g This paragraph will be revised to require the QA organiza:'on to be appointed not less than six months prior to coc=encement of pre ^oerational ~2: ting. 4.4.6 - Test Results Review Qualification require:ents will be addressed for persons responsible for con-ducting and reviewing the results of preoperational and startup testing. 4.4.7 - Training The standard will address qualificaticas for those responsille for empicyee training. 4_.5.1 - Operators The standard will include ccquirenents for unlicensed operators. 4.5.2 - Technicians A high school or equivalent shall be required of all technicians. The standard may address instru=ent technicians separately fran other technicians because of 3reater training and experience requirements. 4.6.1 - Eneineer in Charge This paragraph will be revised to clarify that the engf neer la charge is an offsite position. 5.2 - Training of Porcernel To 3e Licensed by the NRC The folicuing eill be censiderations la rewriting this section. 799 027

M hD J' a l R - e p, p n T, ~ o jlJ -] Provide = ore specific guidance evolutions ~to be performed in the simulator. Provide guidance concerning the content of the training program including suggested nu=ber of hours for each aspect. Provide. guidance concerning acceptable level of performance. At the completion of training at a sNL tor training center, each candidace will be comprehensively examined by the tr nning centar staf 2. on his performance during nor=al and abnormal conditions and certification of the results shall be provided to the owner-ope.ator organization. In addition to evaluation during the training program, the owner-operator organization shall provide a comprehensive examination of each candidace at the completion of training. This ex:3,fnation shall be of the came type and level of difficulty as the NRC-administered examinations. Successful com-plet.lan of this ev' <n' tion shall be required prior to certification of ccm-petency of this individual to the NRC. 5.5 - Operator Retraining and Replacement Training Last cwo sentences of .5.1.3.4 shall be r.oved to the end of 5.5 and first 6 words deleted.

1. ecords indicative of on-the-job proficiency and R

perfor=ance shall be maintained. Rept.ated errors indicativ., of degraded pro-ficiency shall be reviewed by facility management and appropriate training or other corrective actions shall be initiated." 5.5.1 - Recualification Program for Licensed coerators A simulator shall be used to fulfill portions of requalification program. Retraining shall be on an annual basis including present plant preplanned lectures, on-tha-job training, and operator evaluation en a regular and cca-tinuing basis. 2equalification training shall include simulator training in handling single and cultiple e argency and abnor=al conditions. The. training program shall be evaluated by persons other than those directly responsible for the training and for content and quality such as exas diffi-culty and grading quality. Requalificatica training shall include a valk through sufficient in scope to tacure fa:1111arization with the facility. Sinulater portion of requalification training should not be less" than 40 hours per year, recognizing that training is divided betvoen classroes cad board. 5.5.1.1.1 - General (lectires) ~ Y '-'~m nurber of lectures u-ill be identifi d as a requiranent rather than a raccr:endatien. 799 028

(ll@@ ri) W nj o rp n T,; an j 1 t ls Q Lb d b ] J _a 5.5.1.1.2 - Attendsnee Facility =anage=ent shall have all licensed individuals attend every preplanned lecture except as follows. The requalification progra:a shall contain a grade criterion for exemption from attendance at a given lecture. Tha =ini=um grade acceptable for exe=ption from a particular lecture is 30 parcent in that category of the examination. 5.5.1.2.1 - Control Maniculation The following control manipulations and plant evolutions are acceptable for =esting the reactivity control =anipulations required. The starred items sha1.1 he perfor=ed on an annual basis, all other items shall be perfor=ed on a two-year cycle. However, the requalification progra=s shall contain a co==1t=ent that each ladividual shall perf orm or participate in a co=binati,a of reactivity control =anipulations based on the availability of plcnt equip =ent and syste=s. These centrol =anipulations which are not performed at the plant shall be per-formed on a simulator. The use of the technical specifications should be =axi-nized during the simulator control =anipulations. Personnel with SRO licensec are credited with these activities if they direct or evaluate control =anipulations astheyareperfonedgs ?%Q/%'R/HTCR

• (l)

Plant or reactor startupt., to include a rangs that reactivity feedback frem auclear heat addition is noticeable and heatup rate is established (2) plant shutdcun

• (3) Manual centrol of steca generators and/or feedwater 'during startup and shutdcun (4) 2 oration ard/or dilution during pcwer eparatica
• (5) Any significent (>10 p ar:ent) pcuer cheages in 2.uw il rod c:ntrol or cccirculat'.on flav (6) Any reactor pcwer chaage of 10 parcent or greater here load cha1g2 is perfer ed with lead liait coatrol or where flux, tamperature, or speed control is on anual (for HTCR)

L(7) 'Lasa of coolant 1. including significant ?*.R staan generator leaks 2. inside and outside primary contain=ent 3. large and s=all, including leak-r2te deta nination 4. satur:ted reactor ecolant respcnae (?'a) (3) loss of instru=2nt air (if si=ulated plant specific) (9) 1: s of electrical pc-er (and/or desraded p ver ;curces) 799 029

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• (10)

Loss of core caolant flow / natural circulation (11) Loss of condenser vacuu:s (12) Loss of containment integrity, (13) Loss of service water if required for safety (14) Loss of shutdown coaling (15) Locs of component cooling system or cooling to individual components ~ (16) Loss of nornal feedwater or normal feedwater system failure

• (17) Loss of all feedwater (nor=21 and e ergency)

(13) Loss of protective system channel (19) Mispositioned control red or reds (or rod drops) (20) Inability ;o drive centrol rods (21) Conditions requiring use of emergency bor. ion or standby liquid control system (22) Fuel cladding failure or high activity in reactor.ccolant or of Egt.s (23) Turbine or generator trip (24) 1hlfunction of autocatic centrol systcr(s) which affect reactivity (25) Yalfunct!.on of reactor coolant pressuce/volu=e control system. (26,) Reactor trip s (27) }hin steam line break (inside or cutside contairent) (23) Nuc1 car instrumentation failure (s) 5.5.1.2.4 - Review of Abnormal. F=erygv; and Security Frecedures Fire protection precedures. vill be incorporated into this section. The word "sinulated" will be deleted from subites (3). Annual exaninations ccrg,rabl9 in 2 cope and degree of difficulty to an SRC ensainatica chall be gicen to each licensed cperator and senior operator. de 799 030

po D &YS L o w rw x, o 1) N E _ lt l$ _a 5.5.1.3.1 - Annual Examinations The annual written eynnination shall contain categories of examination questions identified in Appendix A to 10 CFR Part 50 plus another category covering ope. rating experiences from similar plants.. A grade of less than 70 percent in any category shall require accelerated requalification in that catepa ry. A grade of less than 75 percent overall requires accelerated requalification in all categories graded less than 75 percent. An annual oral requalification examination shall be given. The examination shall be graded as pass / fail. Persons failing the er'-fnation shall require accelerated requalification. 5.5.1.3.3 - Observation The following rewording is being considered: The program shall provide for syste=atic observation and docu=ented evaluation of an individual's perfor=ance and cc=petency by the i=nediate super-visor. Such observation"and evaluations shall include a review of actions taken, or to be taken, during actual or postulated abnormal and e=ergency conditions. Delete second paragraph. A sanagement evaluation 'shall be conducted quarterly for the syste=atic obser-vation of each shift. The evaluation shall include (as a minicum cverall) shif t conduct, shift turnover, and content of appropriate log bcoks. 5.5.1.3.4 - Accelerated Recualification ?crsons qualifying for accelerated requalifications as a result of annual anam-ination results shall not perform license duties until successfully cocpleting the pregram. Accelerated requalificar'an shall be given in the categories required or areas identified in the orm unination. Successful complation of the program shall be measured by a reexau . tion in individual categories (<70 percent in any category), rep.cating an antire written annual (<75 percent overall) on repeating the oral examination. HJG:SS ~ 7/5/79 ? 799 OM

9 I. ANS-3 2/ ANSI-N18.7-1976, "Anerican National Standard for Administrativu Controls and quality Assurance for the Operational Phase of Nuclear Power Plants" II. This standa d provides requirements and reccc=endations for an administrative controls and quality assurance program necessary to provide assurance that operational phase activities at nuclear power plants are carried out without undue risk to.the health and safety of the public. III. Revise or add appropriate sections: cew7&l A. To stress the en-and and otpashte :tunction of the shift supertisor. B. To clariff control roc = responsibility and the line of ascension in emergencies. C. Provide for 2bhour duty officers in technical disciplines with provisions for i=ediate contact by shift supervisor. D. Define shift turnover responsibilities--each duty statica should have appropriate provision for relieving person to deternine status prior to assn ing shift responsibilities. E. To free the shift supertisor of routine administrative responsibilities and other duties not directly related to plant operation and safety. F. Address rules end authorities governing control roc = access. G. Provide for a secondary enr-and or work station for emergency personnel to be equipped with ec=unications equip =ent and have ready access to reference docunents. The station shall have plant status infornation ~ displays in sene form. 799 032 Ac-1

2 H. Sharpen and strengthen the guidance regarding procedural controls over plant syste=s. Stress that checkoff lists be used = ore extensively. I. Review all NRC positions in Regulatory Guide 133 and detex=ine which should be included in A!E-3.2. IV. Writing assign =ents for the above were =nde to AIG-3 nembers during the last week of June 1979 The r-sults of these assign =ents will receive an initial ec==ittee review on July 31-August 1, 1979 Depending upon the outcome of this review, a revised stnMad could be introduced into the review and approval cycle by August 15, 1779 H. J. Green 7/16/79 e 4 e 9 799 OM

I. ANSI /ANS-3 5-1979, " Nuclear Power Plant Simult, tors for Use in Operator na ni"5" e II. 'lis standad establishes the ~4nd-m require =ents for nuclear power plant simulators for use in operator training and requalificatic,. progra=s. III. The principal =odifiestions to this stnnand will be to revise the list of abnor=al and emergency conditions which the simulator skal' be capable of simulating. Atta+-ant 1 is the currenc listing, and attachment 2 is the proposed revisien. IV. 'lhis revised standard will be reviewed by ANS-3 July 31-August 1,1979, and should be entered into the review and approval cycle i==eddately afterward. H. J. Green 7/16/79 e H O-E 799 034 f.*

n ATTACIIENT 1 M' T D I OQ 3.1.TPinnt Malfunctions. The simulator shall be espable of simulating in real time a O ~D) fd la O yU, '{~-{ g U minirwm nf =cventv.five (79 nbnnrmI_ and emergency conditions resulting from malfunc-tions to demonstrate inherent plant response sad functioning of sutomatic plant controls. Each of the eencrie ccidents anstvzed in tha reference plant safety snsivsis renort which results in observable indications on contro'. room instrumentation sha!! be provided..ni cach shall be considered a single malfunction, The remainder of the minimum number shall consist of a variety of malfunctions associated with the electrical, auxilisry, engineered safety systems and steam systems. Where applicable to the malfunction, the simuistor shs!! provide the capability for the operator to take action to recover the plant or mitigste the consequences, or both. Plant respon 1 to the mr functions t shall be carried out to a reasonable operating condition, as determined by an analysis of the training value of each ms! function. The ab-normal and emergency conditions listed below sha!! be included, as applicable, to the type of reactor. Loss of reactor coolant (Istge and small). Loss of instrument sir. Loss of electrical power (or degraded power sources, or both). Loss of reactor coolant flow. Loss of condenser vacuum. Loss of service (cooling) water. Loss of shutdown cooling. Loss of component cooling (individust com-ponents or Jots! sptem). Loss of fecdwster or feedwater system failure. Loss of neutren ilux indication. Mispositioned control rod or rods (including rod drops). Insbility to drive one or more control rods. Conditions renuirine use of backup reactor _ shutdo3vn,, splams. Fuct cladding failure or high activity in renc-tor coolant or off gas. Turbine trips. Failure cf.utomatic reactivity control

spicms, Steam generator tube leak.

Steam leak (selected sizes) inside and outside containment. Failure of pressure control systems. 799 035 Generster trips. llencf or trips.

ATTACECTI 2 3.1.2 Plant Malfunctions. The simulator shall be capable of simulating in real time a minimum of seventy-five (75) abnormal and emergency conditions resulting from malfunctions to demonstrate _ inherent plant response and functioning of autom. tic plant controls. Each of the generic accidents analyzed in the reference plant safety analysis report which results in observable indication on control room instrumentation shall be provided, and each shall be considered a single malfunction. The remainder of the miniwum number shall consist of a variety of malfunctions associated with the electrical, auxiliary, engineered safety systems and steam systems. Where applicable to the malfunction, tne simulator shall provide the capa-bility for the operator to take action to recover the p!.nt or mitigate the consequences, or both. Plant response to the malfunctions shall be carried l out to a reasonable operating condition, as determined by an analysis of the training value of each malfunction. The abnormal and emergency conditions listed below shall be included, as applicable, to the type of reactor. (1) Plant or reactor startups to include a range that reactivity feedback from nuclear heat addition is noticeable and heatup rate is established. (2) Plant shutdown. (3) Manual control of steam generators or feedwater during startup and shutdown or both. (4) Boration and/or dilution during power operation. (5) Any significant (>10%) power changes in manual rod cor. trol or recirculation flow. (6) Any reactor' power change of 10% or greater where load change is performed with load limit control or wherp flux, temperature, or speed control is on manual (for HTGR). 799 03.6

~ W" u6aS (7) Loss of coolant 'N - ) ~ f[U ~ ~ [, ~ D including significant PWR steam generator leaks a. U O b. inside and outside primary containment large and small, including leak-rate determination c. d. saturated reactor coolant response (PWR) (8) Loss of instrument air (if simulated plant specific) (9) Loss of electrical power (and/or d.! graded power sources) (10) Loss of core coolant flow / natural circulation (11) Loss of condenser vacuum (12) Loss of containment integrity (13) Loss of service water if required for safety (14) Loss of shutdown cooling (15) Loss of component cooling system o cooling to individual components (16) Loss of normal feedwater or nomal feedwater system failure (17) Loss of all feedwater (nomal and emergency) (18) Loss of protective system channel (19) Mispositioned control rod or rods (or rod drops) (20) Inability to drive control rods (21) Conditions requiring use of emergency boration or standby liquid control system (22) Fuel cladding failure or high activity in reactor coolant or o.5 :as (23) Turbine or generator trip (24) Malfunction of attomatic control system (s) which affect reactivity (25) Malfunction of reactor coolant pressure / volume control system (26) Reactor trip (27) Main steam line break (inside or outside containment) (28) Nuclear instrumentation failure (s) 799 037 9

I. FUNCTIONAL REQUIREMENTS FOR INSTRUMENTATION (ANS 4.X) II. SCOPE This standard will contain criteria for deter.aining the functional requirements for instrumentation used to determine plant conditions for nuclear power electric generating stations during normal, abnormal, accident and post-accident conditions. The standard will include both safety and economic considerations. The standard will include guidance on; identifica-tion of expected plant conditions and the environmental conditions associated with these conditions, determination of location and instrumentation type for each application, and determination of type and location of data process-ing capabilities. III. JUSTIFICATION The justification for the standard is to proivde industry, particularly plant designers, with guidance on the selection of instrumentation and environ-mental qualification of this instrumentation for all plant condtions, normal through post-accident. The standard will compleraent other standards, and NRC guidance (e.g., SRP 3.11) but be broader in nature by including economic considera2 ions. Two obvious lessons fror. TMI-2 provide additional justification for this standard. First, the operator must know not on;y what parameters change and how they may be expected to change following any initiating event, but also what parameters are expected to stay the same. All of the instrumentation relied ~upon to monitor these parameters needs to be qualified for the ap-propr. ate initiating events. Second, given the possiblity for the occurrence of unforseen events, the capability for monitoring some plant parameters over extended ranges and for exstensive periods of time needs to be provided. 799 038 Ilo -s

IV. SCHEDULE / MECHANICS It is estimated the standard can be completed in one year with the first draft being produced with three months of project initiation. The working group for this project is being organized now and will have its initial-meeting n/3o 1 ext.cck. The standard will be developed by a working group of ANS-4 in close liaison or conjunction with IEEE (SC-6), with assistance from ANS-51 and 52. Meetinss with EPRI, NRC contractors (e.g., Sandia), nuclear plant suppliers and plant operators will be necessary early in the project to determine instrumentation capabilities and expected plant conditions. The project will commence with a meeting to determine the detailed scope and initiate identification of parameters and plant conditions. Representatives from NSSS designers, A-E's, utilities, IEEE, instrument manufacturers, NRC and EPRI are desirable on the working group. Following the initial meeting writing assignments. will be made. Several meetings in the first few months are antici-pated. Administrative help will be supplied by the member's organization. I 4 9

Sw I. TITLE / ANS - 53.5, PROBABILISTIC RISK ASSESSMENT II. SCOPE PROVIDE CRITERIA FOR PROBABILISTIC RISK ASSESSMENTS THAT DESCRIBE ACCEPTABLE METHODOLOGY, DATA BASE, SENSITIVITY ANALYSES AND APPLI-CATION. THE WORK GROUP WILL ASSESS CRITERIA FOR APPLICATION OF SINGLE, MULTIPLE AND COMMON-MODE FAILURES IN DETERMINISTIC ANALYSES. III. PURPOSE ESTABLISH RATIONAL CRITERIA TO ASSURE CONSISTANT AND ACCEPTABLE USE OF PRA. IV. SCHEDULE COMMITTEE ESTABLISHED CFAIRMAN BEING SOUCHf DRAFT AVAILABLE ' 12/79 ANS BALLOT 6/80 DETERMINISTIC FAILURE CRITERIA DRAFT 12/80 799 040 Ho-7

CRITERIA FOR SAFETY-RELATED OPERATOR ACTIONS ANS 58.8 ~ I. PURPOSE These criteria are intended to provide guidance to the designer to deter-mine when it is acceptable to rely on operator actions to mitigate the con-sequences of design basis events in nuclear power plants. S II. SCOPE Operatt actions following a design basis event in nuclear power plants are required, optional, unplanned, or equipment protective. Required operator actions include those that are part of the plant design basis and which ini-tiate or adjust safety system equipment to provide the minimum acceptable performance. Optional operator actions, are those which are not required by the plant design basis, but may be useful in improving the performance of a safety system; however, the consequences shall be ecceptable in the absence of such actions. Unplanned operator actions are those which may be useful to mitigate the consequences of unforeseen situations. Ecuipment protective actions are those which protect plant equip' ment from damage. These criteria are limited to required operator actions used in the dasign basis events that are examined in Chapter 15 of the SAR fc. pla: t. They establish: Requirements for determining whether a particular safety-related action a. to initiate or adjust a safety system may be accomplished by operator action or shall be accomplished by an automatic protection systen, and b. Functional guidelines for the instrumentation, controls, indicators, etc., necessary to support the required operator actions. 4 799 041 Alo - f

III. BASES AND ASSUMPTIONS These criteria are based on the concept that automatic portection systems will be provided to initiate safety system actions during design basis events where coerator actions do not meet the criteria herein specified. The rules set forth in the following criteria include conservative time intervals and other restrictions to provide an adequate safety margin for the purposes of safety system design and safety analysis of the design basis events. They are not intended to describe expected actual operator action times while responding to design basis events. They are not in-tended to be used in establishing plant procedures or in training plant operators. The criteria of this standard are based upon the following assumptions: a. adequate and centrally located instrument displays are provided as part of the safety system to alert and guide the operator, b. the operators are qualified or licensed in accordance with applicable requirements, c. written emergency procedures are available and are used for periodic training exercises,'and d. the plant staffing requirements assure that an operator will be available to perform any required operator action. IV. SCHEDULE / MECHANICS A. Working group meeting to be held in early ' 4 gust to resolve conments from trial use period and NUPPSCO review. Significant issues to be addressed include:

. 1. Limitation of scope to Chapter 15 analyses is artificial given TMI-2. Realistic analyses of many non-bounding events 911.1 show the need for other operator actions to be evaluated. 2. Elimination of the instrumentation criteria may be possible since other documents may be published which provide more specific guidance. 3. Installction of interlocks should not be such as to preclude an eirly execution of a required operator action if the operator recognizes it needs to be done. B. Werking groups' consultants have met and have agreed to methods of data reduction and analysis to allow them to provide working group with times for the time tests. C. Data collection from simulators has been slowed due to several computer problems and a low flux of trainees. The computer problems have been solved and the flux of trainees has increased. Working group goal is to get enough data by end of 1979 to provide a first draft of the time tests. Data collection'to continue through June 1980 to allow refinement of time tests, and give them a more substantial basis. 4 9 799 043

I. TITLE ANS - 58.9 (FORMERLY N658/ANS - 51.7), SINGLE FAILURE CRITERIA FOR LWR SAFETY RELATED FLUID SYSTEMS II. SCOPE PROVIDE CRITERIA TO DESIGN AGAIWST SINGLE FAILURES IN SAFETY RELATED FLUID SYSTEMS. III. PURPOSE DESCRIBE INTERPRETATION AND APPLICATION OF 10 CFR 50, APPENDIX A REQUIREMENTS, PARTICULARLY PASSIVE SINGLE FAILURES. IV. SCHEDULE ANS BALLOT REVIEW JULY 1979 ANS APPROVAL SEPT 1979 bc:st.fet*Tc kAA [E 5 tafd V A 799 044 No~ f

f REVISION OF ANS - 58.9 1. Change PWR to LWR criteria with appropriate rewording. 2. Revise definition of operator error to be more explicit. 3. Provide additional criteria for dual-purpose systems. 4. Reflect current passive-failure criteria. 5. Provide additional guidance on post-LOCA maintenance. 6. Provide additional criteria for operator mitigation of single failures. 7. Provide additional guidance on passive failure of ventilation ductwork. e 8. The spurious action paragraph of the active failure definition was clarified. 9. The definition of safe shutdown was expanded.

10. Section 3.2 on Condition II events was clarified.

11. A new exemption for having one train of a safety system inoperable due to maintenance in accordance with plant technical specifications was added to Section 4.

4 799 045 j Y' t .- q y>,C...(v

1 ( // I. TITLE ANS - 58.10, REALISTIC METHODS FOR LWR EVENT ANALYSES II. SCOPE ESTABLISHES LWR EVENT ANALYSES METHODS, INCLUDING CONSERVATIVE AND REAL-ISTIC METHODS. III. PURPOSE JUSTIFY REDUCING CONSERVATISMS AND PROVIDE BASIS FOR REVISION OF EVENT ANALYSIS REGULATORY GUIDES, ALARA EVALUATIONS, ENVIRON-I MENTAL EVALUATIONS, MARGIN ASSESSMENT OF ACCIDENTS, AND PREDICT EXPECTED ACCIDENT CONDITIONS. IV. SCHEDULE OUTLINE AVAILABLE SEPT 1979 DRAFT AVAILABLE DEC 1979 ANS BALLOT MARCH 1980 ^ 799 046 /jo-to

SHUTDOWN CRITERIA FOR LWR PLANTS ANS58-i_1 I. PURPOSE This standard provides design criceria for the safety functions which must be capable of being accomplished to ensure the ability to achieve a reactor shutdown from nomal and post-accident conditions. II. SCOPE This standard includesspecific design criteria for those safety functions that are required to perform a safe shutdown of the reactor to the cold shut-down conditions. x0 D The following safety functions will be addressed in this standard: Ak

1. Reactivity Control p

2. RCS Heat Removal

}

3. RCS Integrity 7

a) RCS pressure control ~ b) RCS inventory control i

4. Steam Gcnerator Integrf ty (PWR only) g I

a) S.G. pressure control O Q b) S.G. inventory control i } The standard will include criteria for the nomal and abnormal (i.e. degraded '3 or alternative) systems used to accomplish these functions. u Examples of specific design criteria which will be included are as follows: 9(

1. Acceptability of the utili:ation of non. safety-grade equipment.

2. Single independent failure, redundancy and diversity requirements.

3. Definition of end point of cooldown and time period to accomplish.

4. Safety support system availability requirements.

) 51 MJD Y d.J t.r N.A M D' 799 047 M O -Il

. III. JUSTIFICATION Several incidents at operating plants (e.g. TMI-2, Brown Ferry) have shown 'that current plants are very flexible and include significant alternative means of accomplishing ufety functions even under severely degraded plant conditions. This standard will ensure that in the design phase,the need to ensure this capability is recognized and provided. Draft regulatory Guide '.139, Guidance for Residual Heat Removal was an attempt at addressinC one of the safety functions cited above. This stand-ard will provide a more comprehensive set of criteria for each of the four safety functions cited above. IV. SCHEDULE / MECHANICS A project charter and f 4rit draft outline of this standard will be presented for apporval to ANS NUPPSCO at its next meeting. A chainnan and several members for the writing group have been recruited. As of now, there are no NRC members. This standard has no firm schedule at this time. The working group does not currently have the support for the participation in an accel-g erated development ef fort. It is expected.that this standard could be produced in six months if: 1 Specific requests for its production were received from appropriate bodies (e.g. NRC, AIF), and 2. Personnel were made available from appropriate organizations, and 3. Funds were made available to support its production (i.e. administrative services and personnel time and travel expenses). 9 799 048

I. CONTAINMENT HYDROGEN CONTROL ANS 56.1 II. Present Scope This standard provides guidance for designing systems for light water reactor power plants to assure that hydrogen concentrations which may occur within reactor containment structures following postulated loss of coolant accidents (LOCA) are maintained below levels which might cause deleterious effects on containment integrity or on other required safety features. This standard addresses hydrogen sources and concentration, hydrogen concentration limitations and control methods and specifies control system design and operating requirements. III. It is proposed that the scope of this standard be expanded to include hydrogen generation and control within the reactor pressure vessel (RPV) and containment following transients other than LCCA. The standard will also include calculations (and calculational methods) of hydrogen generation based on temperature and time considerations and calculated detonation limits for resultingh ratios. IV. Schedule and Mechanics for Completing Standard Writing During the period from just prior to the TMI-2 incident and continuing through June, 1979, the ANS 56.1 writing group suffered a loss of personnel, including the NRC representative. Efforts to reactivate the former membership are underway,and recruitment of additional members, particularly those with expertise in core physics and hydrogen generation, is in progress. Particular problems have been encountered in retaining and obtaining personnel with expert capabilities in the area of this standard because they are all heavily involved in TMI-2 related concerns within their own organizations and at present do not have their manage =ent's approval to dilute these efforts to the extent necessary to participate in an expedited standards writing effort. The ANS 56.1 Chairman has requested that, if possible, Mr. D d Green (NRC Containment Systems Branch) be returned to the writing group as the NRC representative. Assuming the writing group is adequately staffed and the scope has been revised and approved, it is estimated that a draft standard can be avail-able for review and comment within six months. / y3 V j]o-it 799 049

I. CONTAIN}ENT ISOLATION PROVISIONS FOR FLUID SYSTFl!S ANS-56.2/N271-1976 II. Scope The primary purposes of this standard are to specify minimum design, testing and maintenance requirements for the isolation of fluid systems which penetrate the primary containment boundary of light water reactors. These fluid systems include piping systems (including instrumentation and control) for all fluids entering or leaving the containment. Electrical systems are not included. The provisions for containment isolation impose additional requirements which are not required for the fluid system function. This standard does not consider any isolation requirements that may exist for controlled leakage areas either enclosing the primary containrant or contiguous to the primary containment. III. It is proposed that the scope of this standard be expanded to include accident isolation criteria, including sequential or preferential isolation and the actuatirg signals requ' ired for each type of isolation and the standard be revised in accordance with the expanded scope. This standard is endorsed, in part, by NRC Reg. Guide 1.141 (April, 1978). Certain exceptions in this endorsement may be able to be resolved by the revision. IV. Schedule and Mechanics for Completing Standard Revision Following approval and publishing of this standard in 19f6, the writing . group chairma n attempted, unsuccessfully, to proceed with a re;ision or supplement on accident isolation criteria. The former 56.2 writing group is now widely dispersed and in many cases individuals are now involved with other activities. Thus, the experts in this area vill have to be located and recruited anew and the working group reconstituted. Because of the impact of the TMI-2 incident on workloads of individuals knowledgeable in the area of this standard it is believed that assistance will be required to, ensure appropriate management approvals for participation in an expedited standards writing effort. Following the restaffing of the ANS 56.2 writing group and approval of the revised scope, it is estimated that a draft of the revised standard can be available for review and comment in approximately six months. 2 799 050 Ho-Q

I. PRESSURE / TEMPERATURE TRANSIENT ANALYSIS FOR LICHT WATER REACTOR CONTAINMENTS ANS-56,4 II. Present Scope This standard provides guidance for the analysis of postulated pressure and temperature transients for containments, including subcompartment and ECCS minimum backpressure analysis. Guidance is presented for the formulation of important input parameters to assure a suitable conservative design. Interfaces with passive and active systems governing the analysis, such as ECCS, structural heat sinks, containment fan coil units, containment spray systems, and pressure suppression systems are identified. III. ANS 56.4 is in the process of finalizing the draft for review and coe. ment. Completion of the section on BWR pressure and temperature transfer analysis has been delayed by difficulties encountered in obtaining appropriate detailed information. Following readjustment of the working group organization and with reasonable response by existing committee members, the finalized draft can be ready for review and comment within three months. 9 }}D- / f 799-051

I. PWR AND BWR CONTAINMENT SPRAY SYSTEM DESIGN CRITERIA ANS 56.5 II. _Present Scope This standard provides the design, performance, testiag, maintenance instrumentation and control requirer.ents for the Containment S , and for boiling water and pressuriaed water reactor stationary electric genera pray System plants. The Containment Spray System, consisting of a Spray Subsystem and an Additive Subsystem, PWR feedwater or PWR steam line break by injecting a water sp containment atmosphere. safety by performing one or more of the following functions:The Contain (1) Containment Post-Accident Pressure Suppression (2) Containment Post-Accident Heat Removal (3) Contain=ent At=osphere Post-Accident Fission Product Removal (4) Post-Accident Mixing of Containment Atmosphere (5) Post-Accident Containment Sump pH Control III. ANS 56.5 is looking into the need for supplementing this standard to address TMI-2 type conditions and should know by August 1, 1979 i~f any changes or additions should be proposed. However, at opinion of the writing group that no changes are required.this time it is the general ^ 799 N //O-l[

e I. ENVIRONMI.NTAL ENVELOPES TO BE CONSIDERED IN SAFETY RELATED EQUIPFENT ANS 56.9 II. Present Scope This standard defines categories of mechanical and electrical equipment an.d ~ the environmental envelopes to which they are to be qualified. There are a limited number of categories of eq:1pment environmental qualification (EEQ) that are characterized both by bounding and normal valuer, r ranges of temperature, pressure, humidity, radiation, and by different ccident scenarioG. These categories cover the complete range of environ = ental conditions anticipated in a nuclear power plant. The ECQ envelopes shall be used for equipment qualification. III. ANS 56.9 has completed a major rewrite of the standard and is reviewing the rewrite in light of com=ents received from NUPPSCO and TMI-2 considerations. Preser/cly, progress is being retarded by other workloads on the working groep chairman. However, it is estimated that a final draic can be available far NUPPSCO ballot in November, 1979. h 9 799 053 MC-/6 .}}