ML20052D426
| ML20052D426 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/27/1979 |
| From: | Mattson R NRC - NRC THREE MILE ISLAND TASK FORCE |
| To: | NRC - NRC THREE MILE ISLAND TASK FORCE |
| Shared Package | |
| ML19257E308 | List: |
| References | |
| FOIA-82-92, RTR-NUREG-0585, RTR-NUREG-585 NUDOCS 8205060486 | |
| Download: ML20052D426 (2) | |
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DEC 21 G3
-l MEMORANDUM FOR:. THI-2 Action Plan Task Managers FROM:
Roger J. Mattson THI Action Plan Steering Group
SUBJECT:
AIF REVIEW 0F NUREG-0585 Attiched for your information is a copy of a review of the "TMI-2 Lessons Learned Task Force Final Report" NUREG-0585, done by the AIF. The Steering Group has not identified any coment that requires substattive change to an action plan. Some clarification of actions may be appropriate
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6as~e'd 6h' the'se comments.
s Originals!gned by Roger J. Mat + Ion P
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Roger J. Mattson
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TMI Action Plan Steering Group
Enclosure:
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December 27, 1979 MEMORANDUM FOR~:- TMI-2 Action Plan Task Managers FROM:
Roger J. Mattson
-7 TMI Action Plan Steering Group
SUBJECT:
AIF REVIEW 0F NUREG-0585 Attached for your information is a copy of a review of the "TMI-2 Lessons Learned Task Force Final Report" NUREG-0585, done by the AIF. The Steering Grou'p has not identified any comment that requires substantive change to an action plan. Some clarification of actions may be appropriate based on thsse comments.
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Roger J. M&ttson TMI Action Plan Steering Group
Enclosure:
Ltr dtd 12/13/79 fm Byron Lee, Jr.
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December 13, 1979 i
Mr. Harold Denton Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
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20555
Dear Mr. Denton:
Several subcommittees of our Policy Committee on Follow-Up to the Three Mile Island Accident have reviewed the "TMI-2 Less~ons Learned Task Force Final Report" (NUREG-0585).
The resu-Its of this review, as endorsed by the Policy Committee, are encias.ed.for your attention.
In general, we agree with the principal thesis of the report that operational safety factors require proportion-ate-ly greater attention.
The efforts of our subcommittees and the initiation of the Institute of Nuclear Power Operations have already set in motion many of the remedies.
We believe
-great.. car 6 should be taken on your parx to develop new regu-latory guidance which encourages the industry momentum already established and avoids prematurely, implementing programs that either, duplicate effort or establisk hard and fast fixes tq_a every perceived concern.
Certainly one of the lessons learned from TMI is the need to improve communications between designers and operators to be assured that actions. in one area don't -
have negative reactions on the other.
riith the concept of an established safety goal..and operational' safety measures already"~t'aken and proposed to
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complement current design margins, the report offers a model for constructing pathways to a future regulatory system.
The evolution of this system must proceed in a timely, but orderly, manner in parallel with the short-term programs and should encourage, not constrain, continued plant operation and resump-tion of operating and construction permit licensing.
We agree with the intent of most of the actions recom-mended in the report.
Our concerns with certain sections are stated in the enclosed responses.
We continue to have concern about schedules and priorities for the multitude of actions being proposed.
We offer our assistance in resolving any of these concerns.
Sincerely, 7
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Byron' Lee, Jr.
Chairman, AIF Policy Committee on Follow-Up to the TMI Accident
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C:: TH E T 2 ::.' ~_.'. C : :.".E: T AT ~. C ::S C 7 THE T::I-2 LESSO::S LEAF.::ZD TASK' FDF.CE PERSONNEL QUALIFICATIONS AfiO. TRAINING 1.
1.1 Utility Manacement Involvement definitive presence
. The corporate management o'f each licensee should establish a and involvement in the'sslection, training, and qualification of operations To assure that th'is has been accomplished, the NRC should require, personnel.
as part of the app,lication for operator and senior operator licenses, that corporate management certify.the competence..and fitness of the applicants.
Such certification should'be required by the highest level of corporate manage-cent r.esponsible for plant operation (for exa=ple, the Vice-President for s the'
. Operations)...The Task Force recommends that, when the dRC staff judge.
quality of applications from a particular utility to be deficient, the corporate official certifying the competence of the applicants be required to discuss the reasons for the decline in ccmpetence and planned corrective action with the Director of Nuclear Reactor Regulation;.
' The involvement.of corporate management of each licensee is
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competence and fitness of applicants _for, operator and' senior __
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operator licenses is a responsibility of the corporate managemeEt responsible for plant operation.
For this reason the subcommittee concur's that the responsible senior corporate management be held accountable for the competence i
as well as operator perfor.
of-' operator lice nse applicants,
mance.
In carrying out this responsibility, however, corporate
, management should have the flexibility to delegate the certifi-cation function to others in the organization.
In the vieth of l
the subcommittee, the plant manager,, who is charged with the safe operation of the plant, is the lagical person to certify the com-j._
petence and fitness of the applicants.
The role of corporate i
l management is to assure that the licensee's system whicn selects, t
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trains and qualifies operations personnel, provides quality applicants.
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i 1.2.Traininc procr:ms Each licensee should be required to review, within one -year, its training prcgram for all.cperations personnel, including maintenance and technical personnel, and
'should justify the ' acceptability, of training piograms on the basis that tnese programs provide sufficient assurance that safety-related functions will be effectively carried out.
Documentation of this Feview and justification should
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be retained on site for inspection, b'ut need' not be submitted to the flRC for m e -" -'.
r eview~. 'The.p' referred method.of fulfilling this recommendation is~ a' position task analysi's, in which the tasks performed by the person in each position are,,
dhfined and the trainingt in.Aonjunction with education and experience, is
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identRiid to previde assurance that the tasks can be effectively carried out.
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The position task analysis should include normal and emergency duties, including maintenance activities, placing amphasis on..the role played by~every member cf an operations organization in assuring safe plant opert.tions.
All levels of..-
- the operations organization should be includeddhis action is regarded by the Task Force.as-an. interim measure pending resolution of the question of licensing of additional operations ' personnel beyond reactor operators and senior reactor operators, as discussed in Recommendation 1.8 of this appendix 1
- The scope of emergency duties defined in the positionTask analysis should not
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be restricted to 6nly the transients and accidents ' considered in the design
_ basis. 'The tr,ainiiig should. recognize that events beyond the current licensing des 1gn~5 asis events can occur.
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~ystems already installed 'at the
. The training should include the use of them s plant to control or mitigate the consequences df accidents in which tiet core is. severely damaged.
This training would be an interim measure pendirig com-pletion of the rulemaking to determine what design features to mitigate the.s.e more severe accidents should'be required..
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~~ The'need to ve'rify the acceptability of training progfams, is recognized and is a primary function of the industry's-newly created Institution of Nuclear power Operations.
In the interim period, while the Institute is being formed, each licensee.is evaluating and upgrading its training program in light of the lessons of TMI.
With this recog-nition, the subcommittee concurs with the Lessons Learned Task Force recommendation that each licensee conduct a revie'.Tof its' training program.
The subcommittee bel,ieves it should aphly to only the operations personnel, including maintenance and technical personnel assigned to the plant site.
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- l. 3 -In-Plant Orills
.. z Each licensee should be requj. red to review, within 90 days, its training For tasks performed
'progra.r with respect to the conduct of in plant drills.
by shift operating perscnnel in res;:ense to off-normal or accident situati'ons, licensees should assure that sufficient in plant drills are. conducted to enable ~ personnel to main,tain proficiency in'those tasks.. The Task Force considers drills of a walk-through nature acceptable 'and does not mean to imply the_ actual manipulation of controls or equipment or initiation of an event (such as by the opening or closing of valves o'r tripping breakbrs or The Task Force considers that drills requiring the physical manipula-pumps).
tion of cont?ols are also important but can be more efficiently and safely conducted using an appropriate nuclear power plant simulator.
With this in This mind, each licansee should deTelop a schedule for ir. plant drills.
schedule should be a part of a disciplined training program for each station.
- It-n'ee4.no.t b~s submitted to the NRC for review; however, it should be avail'able
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t at the site for inspection.
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The subcc::nittee fully supports the recohmendation for in-plan't dri-11s.
It is noted that the Institute for. Nuclear
- Power Operations will provide benchmarks for the conduct and frequency'of in-plant drills.
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- 1. 4 0cerator Licensina The first areas of personnel qualification that need to be upgraded are those NRR pertaining to licensed senior reactor operators and reactor operators.
recomendations to the Commission for improvements in the operator licensing We believe program were contained in Commission Paper SECY 79-330E (Ref. 2).
l these reco:cendations should be treated as the first steps in a long-term They are, however, necessary improve-program to upgrade operator proficiency.The ultimats resolution of the issue of qualification cents in the program.
of reactor operators should take a broader perspective.
Although the Task Force generally agrees with the reco==endations contained in.SECY 79-330E, we impl~erentation of the following additional items by the regulatory reco=end staf f in conjunction with the icplementat-i.sn of t.he rere:reendatinr. :
n SECY 7S-330E.
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i (1)' As part of the inspector tr.ining program of the Of fice of. nspection and I
Enforcement (IE), operater licensing program personnel of the.7ffice of
,.., Nuclear Reactor Regulation should (a) provide information to IE inspectors ori th'e operator licensing program and- (b) iBentify the types of information the IE inspectors should provide to assist flRR in rnking decisions with regard to the renewal of operator licenses.
-Th'e subcommittee concurs with the Task Force recommendation
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. to improve commtinicatior.s and interaction between the
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Office of Inspection and Enforcement and the Office of Nuclear Reactor. Regulation.
It helieves implementation
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.of this recommendation will help to improve licensee respense.
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The NRC staff should establish a cechanism whereby individuals..cormitting (2)
Such a operational errors are identified in Licensee Event Reports.
. c.,pechanism should include provisions for protection of the pr.ivacy of the L
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The intent of this recoc6ndation is to: provide atditional infor=ation to operator ' licensing program personnel to assist them in determining the continued qualification of operators in the review of operator license renewal applications.
Due consideration shoul?-be given,
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- to whether such reporting will affect the quality of reports received by t.he RRC.
The subcommittee believes that privacy laws will make
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itdifhicult, if not in possible, to exercise fully the
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potential of Licensee Event Reports to improve opera-tional safety if individuals committing operational
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errors are identified.
'Suen a requirement coulc have the negative effect of operators attempting to cover-up 1
mistakes to avoid unfavorable.public identification.
How-ever, it is the subcommittee's view that such information L-should be maintained by the licensee and made available l
l toNRChersonnelat the plant site.
The recommendation'to require the identification in LER's of individuals com-mitting operatienal errors should be reconsiderec.
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- e r-As part of the trainih; progr:m for all licensed : erst:r:,
should_be cencucted by the tiRR c;cra.cr li:ersing 'pr: gram p{,p nel
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I cour:2 wi'th assistance fr:m other tiRR technical personnel.
Particuiars o.
.e course would include
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(a)
S$fetyanalyses (b)
Prebabilistic assessr.ents (c)
Current safety issues and recent significant operating experience (d). t4RC and industry responsibilities for safety This recercendation woUld.'r.einforce the knowledce of and respect for.
acc'icent/ transient sequences as well as providir.g a pcsitive feedback for
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better decisions by tiRC staff on reactor operations and design matters.
Additional t1RC staffing will be required to accomplish this objective.
s This is one of a number of Task Force reccamendations
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which would provide an operational feedback to NRC per-sonnel.
The subccmmittee. believes that the feedback function can be..better. served through increased inter-
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subcommittee believes that the responsibility for the training of licensed operators is a function of the 1).censee and does not consider it prudent to weaken this recponsibility by having NRC undertake a part of th'e licensed operator training program.
Prior to assuming initial' assignment as shift supervisor or shif t technical (4) advisor and on a biennial basis thereafter, individuals should be interviewed by an interdisciplinary group of fiRC staff.
Such interviews should probe the individual's technical knowledge in the area of transient and accident
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in the case of a shif t supervisor, the managerial ability respense and, to cca:and and control th'c activitics of shift personnel.
These interviews should be coriducted at tiRC headquartars.
Criteria for
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subjects to be covered and acc~cptable standards of performance o.f individuals steuld be developed by ;;RR operator licensing persennel prior to pr =ulgatien of this recuirement.
This action will require a consideracle expenditure of rescurces and its phasing neecs to.be carefully cgnsidered.
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The subec=ittee finds that this Task Force recc=enda-tion would weaken the licensee management responsibility for the qualification and-assignment of shift super.visory and technical personnel.
The recommendation would also
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_ requh.re a considerable expenditure of NRC and licensee resourc s without a-cc:r.mensurate increase in operational safety.
An appropriate function of the NRC is to approve requirements developed by INPO for shift supervisors and shi t technical advisors and to audit the actions of the licensee to meet the requirements.
The' NRR. operator licensing program personnel should sponsor an annual,
(5) wor.kshop for licensed operators to be attended by at.least one represent-ative of the licensed shif t personnel at each. unit.
The purpose of this'
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.c woFKshop is to provide..an opportunity for_ exchange of information pn
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operating experiences between the NRC staff and the utility shift personnel.
For example, such a seminar could lead to. an' exchance of informatibn on (a) NRC safety concerns related to shift opeirations7 (b) the impacQf
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licensing on shift activities apd.per'sonnel, and (5). i eccamendaticirs frca
. shift personnel concerning changes 'in'ri'. actor regu'la' tion that would '
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improve safety.
The subcommittee believes this rec 6minendation-has '
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merit in providing the necessary operator feedback to NRC operator licensing program personnel.
These
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annual workshops, coupled with improved communications and interchange between NRR and I&E should be instru-menta1 in fulfilling the necessary feedback function.
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As a less prescriptive alternative to Rec:.- endati n 5 of SECY 79-33CE g
that "" Phase II, III, and IV cold training ' program instructors and all h0L (6) tra'ining program instructors that provide instrbcticn in nuclear pcwer pl~ ant operations hold senior. operator licenses and be required to success-fully participate in' app?icable requalification programs to. maintainSuch their instructor status,'.' the following is considered acceptable:
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instructors.should hold or have previously held a senior reactor operator a cc parable nuclear power plant and currently possess (SRO) license or) i
_. instr 0ctor' certification from the Institute of Nuclear Power Operat ons, -
provided the INPO certification program has beer, enmined and found -
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' C acceptable to the NRC.'- Er;hasis should be placed on an instructor's abili't to instruct, in addition to his technical cc petence.
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The~subecmmittee agrees with the Task Force recommen-da. tion.that f.he indicated training program instructors hold or have previcusly held a senior reactor operator license on a. comparable nuclear power plant.
The sub-
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cc=mittee is pleased that the instructor certification function of INPO is recognized and will help to insure y.
that INPO warrants this confidence ? 5. cme clarification
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' is.needed, however, in the Task Force recommendation.
For example, it is not believed th'at health physics and other t.echnical'succort instructors'were intended to meet the license requirement.
It is also noted Ehat individuals holding an SRO license are expected to be in short supply over the next three years, which may require'some flexibility in implementing this recommendation.
Upgraded requirements for acquiring
- i and maintaining an SRO license, and the increased I
demands of vendor and service organization training
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programs in addition to l'icensee operating and training I
dumands will all add to tne shortago.
It is suggested that this TasP. Force recc.mondation be broadened to include trai.ing program inscructors who hold or have previously, 4
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c.2 rsspens.rility at a Copar:- an cf 22fsn:- faci icy.
Consiceration should.be given to placing resicent operator licensing exami5ers in each of the 16ajor geographical areas in whica there is a (7) concentration of training centers using nuclear po,.er plant simulators.
The intent of this recommendation is to provide for greater interaction by operator licensing examiner's in operator, qualification and requalification
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programs.
._-The subcommittee concurs with this Task Force recommen-z- - r dation and believes it.will 51so assist in fulfilling the feedback function for NRC operator licensing program I
personnel.
- 1. 5 NRC Staff Coordination At the present time, several groups are addressing.the subject of qualifications Even though each of the
- of personnel somewhat independently of one another.
- ei. fonts..is, appropriate on a short-term basis, a coordinated approach cust be developed for the long term.
The NRC should9ncrease the staff restmrces in this area, assure the hiring of needed professional disciplines to increase'
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present staff capabilities,,and designate-* responsibilities and organizational -.
S-entities within the various offices.
The subcommittee notes that INPO will provide industry
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coordination on the subje ' V o f qualifications of personnel.
- 1. 6 Licensed Goerator.iuall i: e.itions a_
A program for raising the qualification requirements for shift supervisors and
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senior reactor operators should be established.
The distinction being made in present practice between senior reactor operators (e.g., shift foreman in a multi-unit station) and shif t supervisors should be recognized.
As a short-term action pursuant to NUREG-0578 (until such time as staffing and qualification of shift personnel and the. control room man-mach.1e interface requirements are upgraded), each licensee has been required to provide an on-shift technical advisor to the, shift supervisor.
Within the next five pears, it is recommended that the qualifications of senior reactor operators and shift superv1 ors be QI;alification requirements for-applicants for upgraced as indicated below.
licensing prior to initial fuel loading may require special additional cor.sidera -
tions, partic'ularly with respect to experience.
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- C i3 committed to raising the qualification requirements for operating personnel.
Shift Supervisor (person in charge of operations on shift at the station) ~
(1)
- -- Shift.Sepervisors should have at least a Bachelor of Science degree or equivalent training and experience in engineering or the related physical The Shift Scpatvisor should also hold a senior reactor operater's
-sciences.
license (issued under new proposed requirements defined belcw) and have
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served as a reactor operator for one year or senior reactor operator for six r5cnths.
In establishing equivalency with a Bachelor of Science degree, consideration._ should be given not only to formal courses in-engineering and related sciences, but also to education in-the liberal
-It i~s recommended that the use of the equivalency to a Bachelor of
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Science degree be exercised to only a limited degree and that most shift supervisors hold degrees.
It is also recc : ended that shift ' supervisor
. qualifications include leadership training and efperience.
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The regi2irement for the shif t supervisor to have at
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least a Bachelor of Science degree or' equivalent training and experience in engineering or the related physical c
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(The Task Force recommends that the equivalency'be
. sci,ence exercised only to a limited degree and that most shif t supervisors hold degreas.) is. believed to have.only a short term appeal, and may create instability in the longer term.
Two f actors which must be analyzed carefully before this recommendation is implemented are:
A degree requirement for shift supervisors would o
block the current career path of other operators i
increasing turnover and' recruiting problems which again would create instabilities in operations.
Experience.has shown that keeping degreed en ineers o
on shift-work is difficult and leads to increased turnover in operating personnel.
The increased i
I, turnover creates instabilities in operations.
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.. :.G E::perience has shcwn that an engineering degree does not
. i.{sure developing a competent operator or that the lack
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operators.
Therefore, the subcommittee reccamends.that this Task Force recommendation be evaluated against e: ~
As an alternativ,
possibleinstabiligesinoperations.
a dedicated education and training program could provide the same end result as an engineering degree wiEhout d~reating new op rational instabilities.
(2) ' Senior Reactor Operator (e.g., shif t foreman in.a multi-unit station) -
Senior Reactor Operators should have at least the same general tech:;.ci.
education and.. specific training in transient 'and accident response charac-
.. teristics of nuclear power plants as recently articulated for the shift Additional recommeedations for upgrading sanier rear. tor
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" 1.schhical advisor.
operator qualifications are identified in the Commission Paper SECY 79-300E on Qualification of Reactor Operator-s.
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'The. comment indicated in (1) above applies.
a basic fundamentals cour_se of approximately.. twelve weeks is (3) At presen3.,
required as part of the operator training program.
A prerequisite to satisfactory performance of nuclear power ope *ation is the fundameqj:al The Tase. Force beieves twelve weeks understanding of nuclear technology.
to be insuf ficient time to provide a broad and comprehensive level of It is recommended understanding in the fundamen.als of nuclear ter$nology.
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that the NRC, perhapsI~in consultation with INPO, examine the content of the basic fundamentals course and establish definitive instructional requirements for the course.
The subcommittee agrees with the Task Force recom=cndation that the content of the basic fundamentals course be c::amined and de finitive. ins tructional receirenents be
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estabikshed.
As the Task Forco recogni::cd, this is a primary function of I::po.
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-e The subccrr.ittee notes, hcwever, that the. length and recuirements 1
of the basic course are dependent upon the entrance requirements of, For example,'the course length would differ those taki,:ig the course.
-if the trainees have significant experience in another type plant or hold a. degree. in nuclear engineering, than if their experience and
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Thus, classrcem hours as a Requirements based on course mhteria'l
- criterion should be avoided.
should -be established and reviewed to insure operator qualification.
o 1:7 Licensee Ye~chriical and Manacement Su:: cort The review and evaluation (being co'nducted by the Quality Assurance Branch) of
'the canagement and te,chnical resources available to utilities who own and operate nuclear pcwer plants to-handle unusual events or accidents should be cecpieted, and regulatory guidance should be developed that covers the capa-
- b-ilitiesraad, role of technical and management personnel in the normal operation
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The criferia should contain a requirement o.f the plant and-during an energency.
for periodic verification of the licensee's technical and management support The present criteria for *
. capability throughout the operating life of"the giant.
determining the acceptability of licensee technical and managecent supppt is
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very. general and applies only to. normal pl, ant operations.
i The subcommittee agrees with the int'eU. ~of this Task F6Ece "
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recc'.9bendation and notes that high priority is being given for INPO to. develop the criteria for technical and manage-1 for both normal plant operations and during an f
ment support NRC should review the criteria developed by INFO t
emergency.
before developing any new requirements.
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Licensine of AdditicEal'Oceratinc Ferronr$el
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- 1. 8 The staff should decide which plant perscnnel, other than. react:r cperaters and senior. reactor operators, should be licensed.
NRC review of the, training
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-aiad qualifications of nonlicensed personnel has been very limited in the past,
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based on'the assumption that it is the licensed operators who have the most important influence on plant safety.
A numbe'r of examples from the.TMI-2 accident indicate the degree to which plant safety can be greatly influenced
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1 by per. sonsin dany, positions, including managers, engineers, auxiliary
- operators, maintenahce p'e?sonnel and technicians.
All of these previously nonlicensed 1__~ pe,rsennel may. affect plant operation, and their roles sticuld receive greater Answering the questions of h w much attentic'n'ficm a safety perspective.
independent examination of their qualifications and training is necessary and
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The pre-whether NRC licensing is appropriate is a significant undertaking.
requisites to an effective examination program are definitive qualificction
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The current NRC guidel_ines requirements and specific training programs.
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addressing no.dicensed personnel training and qualification are very general and are not suitable for a' licensing program.
-The newly force'd Ins,titute of Nuclear Power Operations intends to develop standardized training requirements for technicians and nonlicensed operators -
The Task
'and to provide certification for the training of 'these personnel.
i Force believes this program, if properly implemented in a timely way, could We-for ~ detailed guidance frcm NRC, and_cculd, under the right conditions,
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be endorsed by NP.C as meeting its independent licensing requirements fBr addi-tional operating personnel.
A statement of underst'anding between INPO.and the NRC should.be established at an early date Twithin the next six months) so that
. both groups can decide whether and to what extent to proceed independer$4y.
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."Again the subcommittee is pleased that the Task Force recog-ni::es the role of INPO in developing' slia'ndardized training i
requirements for technicians and non-licensed operators and in the certification of training programs.
It is noted that
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for the interim period while INPO is being established, the-sub-l committee has started a program to develop benchmarks for f
the training and qualification of ItC technicians and non-t I
licensed plant operators.
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i o The Cc-.ission's regulations should be revised to more clearly state present staff recuire.ments (as described in the Standard Review Plan, Secti:n 13.1.2)
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foi minime= shif t staf fing of licerised r'eactor operators.
The governing regulation,10 CFR 50.54(k), states that "an operator or senior operator licensed pursuant to Part 55 of this chapter shall be present r.t the controls at all times. during operation of the-facility."
For single-unit power stations,
licensed senior-e-
- q.the_ s.taf f ~ requires _ the shift crew to include at' least one reactor operator, two licensed reactor operators, and two additional operators For cultiple-unit power (auxiliary operators) during, reactor operation.
stationi Gith separata control rocas, the staff also requires the shift crew-
~_ _ '
~ to include at least one licensed senior reactor operator and two, licensed
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"For cultiple-unit power stations reacter operaters for each operating reactor.
with a c6:=on control room, the staff permits a reduction of licensed reacter operators to one per unit plus one additional reactor dperator with the other requirements. rjmaini.ng the same.
However, the staff does not require the presence in the control room at' all times of two licensed operators and the In developing the revision to the regulations, senior reactor operator.
consiceration should be given to requiring the presence -in the control roem at all times during normal operations of two reactor operators and one senior Provisions for tours of the plant by operators will probably reactor operator.
need to be made if this staffing proposal is adopted.
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The subcommittee agrees with the requirement for having at least one
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licensed senior reactor operator and tw& 14censec reactor op. era _ tors both
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in single-unit control rooms and in common control rooms for mT21tiple units.' It is considered important that provisions be included..for tours of the plant.
Plant tours by the licensed reactor operators add to the safe operation of the plant through identification of potential problems,
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and the tours increase the knowledge aid experience ofthe operators.
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3.
WORKiliG HOURS Each licensee'should be required to review and revise wi' thin 90 days the plant
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administrative procedures to assure that a sound policy is established covering It is working hours for reactor operators and senior reactor operators.
recognized that this is a complex subject involving other interests (e.g.,
labor unions).
The tiRC staff should assure that the subject is addressed in a comprehensive manner by all licensees and that the other interests not be allowed to interfere with the-basic safety' interest.
As general guid: ace, it is expected that licensees' administrative procedures will ' cake it unlikely that personnel would have to be used for more than two consecutive wurk perieds in excess of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and that a 12-hour rest period would be required between work period:.
In the event that special cir:umstan cs arise that would cause extenced periods of work in excess of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for more than two c:nsecutive dJys, such work should be auth0ri;cc ty the Sta*,i:n.$nacer with a?pT ?ria*e d:cumentation of the cause.-
Indicati:ns aside fr m Three Mile Islanc lead the Task forte to concluce that this stet cust be taken :: ressena:!y assure that incivicuais are in proper pnysical c:nditi n to perf rm wcrk at nuclear ;:-er
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The subco==ittee agrees with the Task Force recc=er.dr.:i$n.
also recognizes the complexity of establishing working It hhurs i'n'clants with unionized work fo'rces and believes flexibility in implementation may be necessary.
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EMERGENCY PROCEDURES Emer.gency. c'perating procedures for all nuclear power plants should be rev
-~ v The review should be conducted by interdisciplinary review groups by the NRC.
corprising I&E inspectors and NRR technical reviewers knowledgeable in system design, accident analysis, operator training, theories of education and crisis human f acto'rs, and the underlying technical bases for licensing.
Special attention should be paid to the recent advice of the ACRS on the style management, s'
A-safety esialuation regarding the adequacy and tontent of emergency procedures.
of the emergency precedures should be issue'd at the conclusion of the review.
Previcus NRR reviews and' I&E reviews.cf emergency operating procedures did not specifically investigate their compatibility with the' design bases of the systems involved nor was the discipline of human factors included. '
-.Thisaction vill require a censiderable expenditure of resources and its
-It may be satisfactory I.a limit the i
phasing n'eeds to be carefully considered.
l general application of this recommendation to new operating licenses for the These initial few reviews by the staff, with oversight by next year or so.
the ACRS, 'will provide the t'ime and experiendD necessary for the steff and.
- industry to develop and agree upon acceptance criteria for the deveitpcent, formattin'g, and future review of all emergency operating procedures. ' Upon cc pletion of these acceptance cri.t.eria,'say within the next two years, a systematic effort 69 all licensees to review their emergency. procedures and
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ravise them as necessary could be conducted more productively than it could
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today.
While the subcommittee agrees with the Task Force reco=en-dation, it notes that the interdisciplinary review groups do not include operator experience.
It is recommended that the interdisciplinary review groups include personnel with
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current operating e::pcrience '.in the plant type for which the' emergency procedures are being reviewed, and that in the review, the' emergency procedures be walked through at the plant.
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G VERL:IC-[kCN OF CCkRECT PERFORy.ANCE OF OPERATING ACTIVITIES q
b.
A mor.e effective system of verifying the correct performance of cperating activities is needed to provide a means of reducing human errors and improving the quality of normal operations, thereby reducing the frequency of occurrenceSuch a verifica-2 of cituations that could result in or contribute to accidents.
tion system should include automatic system status monitoring and human verifica-tion of operations and maintenance activities independent of the people performing 1
l the adtivityk The Task Rorde recommends that automatic status monitoring be required by a decision to backfit Regulatory Guide 1.47, " Bypassed and Inoperable Status" Indication for Nuclear Power Plant Safety Systems," to plants not, already Furthermore, the design to satisfy the object.ives of the required to meet it.
guide should be flexible and capable of accepting ' additional monitoring functions -
at a later date.
The implementation of Regulatory Guide 1,.47, although reducing the extent of human verification of op~erations and maintenance activities, does not eliminate the need for such verification in all instances...TherTfore, each licensee surveillanhe "should be required to review his procedures for maintenance, test, and other normal plant. operations activities (1) to delineate each activity
_.. _that;te, quires-. independent verification because of its importance to safety, y
(2) to ide6tify the personnel. responsible fote conducting the verification, and
-(3) to describe the method of documenting performance of the verification The results of this work should-b_q. submitted to NRC within 'six months for Use in the development of minimum acceptance criteria for operations process.
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-The verification procedures, probably in the form of a Regulatory Guide.
procedbres' adopted by the licensees should contain two phases; namely,'before and af ter instaliation of status monitorin'g ecuipment in conformance with
The subco: mittee agrees that adequate assurance is needed to verify corfect performance of operating activities.
It does not believe that backfitting Regulation Guide 1.47 is necessarily the best
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method for each plant to achieve this objective.
There are other methods, both human and mechanical, which can provide effective verification.
For this reason the subcommittee believes that licensees should have flexibility in achieving verification.
The review requirement is supported by the subcommittee, but with a word of caution.
A six-month time requirement for licensees' review is another near-term task which must be. accomplished by the same indi-viduals already overtaxed to meet other requirements, many of which are believed to be of higher priority than establishing a verifi-cation guide.
It is suggested that the NRC consider a generic review which could be accomplished on a selective basis using the capabilities of qualified contractors / vendors.
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EVdUATION OF OPERATING EXPERIENCE 6.1 Natienwide Network
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An integrated NRC-utility program to evaluate operating experi'ence should be Action within the NRC has been initiated to establish an Office i
l df Operational Data Analysis and Evaluation to provide agency wide coord'ir.at established.
and,,an overview of all cperational data analysis-related activities performed i
i The nuclear industry, through HEAC and INPO,.
l eq within the 1-ine offices of NRC.
Pursuant to the has established its own eperational evaluation program. licensee is now reouired to have al
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recc.-i::ier.dations of NUREG-0578, eact The director of the new NRC Office of tiens experience evaluation group.
Operational Data Analysis and Evalua icn should take the lead to assure tha.
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these diverse programs are formally tied together to the extent necessary to benefit-frca one anothers' viewpoint and analysis while reccgnizing their,
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i nd i v_i du ai. r.e_s_po ns i b i l i ti e s.
The subcom:fittee recognizes the diverse efforts underway in the evaluation of $perating experience and ac;rees that a cooperative
. effor,t,is desirable.
It is noted that the evaluation of operating experience is a major function of INPO and NSAC.
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c 6.2. Providino Information to the Ocerator Each licensee should be required to review, within oO days, its administrat'ive
-procedures to assure that a mechanism exists through whi Circulars.and Motices, and applicable Licensee Event Reports) and frcm the f
the licensee's own operating experience evaluaticn group are ccnveyed to reactor operators and other affected operations personnel.
Two ways of accceplishing this objective are (1) standard distribution lists.
or publications and (2) regularly' scheduled lectures as part of operations This recommendation is intended to assure that operators staff retraining.
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and other operations personnel are continually provided with lessons learned frcm opcrating experience.
The subcommittee agrees with providing information to the operator and believes that each licensee is already taking action to provide operators with t'he lessons Icarned from operating experience.-
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SU3 COMMITTEE ON CONTROL ROOM CONSIDERL.TIO::S CI:'_**ENTS ON THE FINAL RECOMMENDATIONS OF-THE TMI-2 LESSONS LEARNED TASK FORCE MAN-MACHINE INTERFACE 7.
7.1 Control Room Reviews
~All 1-icensee's should be required to conduct a one-year review of their control rooms.
The safety review should consider control room design
-- =and control room operational procedures, including emergency operating procedures.
In this review, the licensees should evaluate:
(1)
The adequacy of information presented to the operator to reflect plant status for normal operation, anticipated operational occurrences, and accident condtions; (2)
The grouping of displays and the layout of panels; (3)
Improvements in the safety monitoring and human factors enhancement of controls and control displays;
. _. (4)....The communication from the control room to points outside the contrdi' room, such as the on-site Technical Support Center.
(This communications link must also be coordinated with new requirements for transmission of plant systems data to NRC.);
c (5)
The use of direct rather than derived signals for the prelentation of' process and safety information to the operator;
,(6)
The operability of the plant from the control room with multiple failures of non-safety grade and non-seismic systems and control room syste'ms; (7)
The adequacy of operating procedures and operator training with
" respect to limitations of instrumentation displays in the control room;
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(8)
The categorization of alarms, with unique definition of safety alarms; and (9)
The modification of operating procedures and operator training programs as a function of control room modivications resulting from this review.
The purpose of this recommendation is to improve upon operator process communications.
Guidelines and criteria for the centrol room design review are now being drafted by the Division of Systems Safety, including consideration of the results of previous studies of this sort and existing technology outside of the nuclear industry.
Explicit criteria can probably be developed by about February 1,1980.
Consideration is being given to a series of topical meetings with recognized experts in the field and affected licensees.
Specific 8
requirements for backfitting existing control rooms to correct deficiencies will be established in the course of the reviews by licensees.-
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This section is in basic. agreement with the Sep ember 19, 1979 report of the AIF Subcommittee on Control Room Considerations.
Further refinement and definition of the NRC recommendation is needed.
The last paragraph of this section stated that guidelines
- r- '1 and " criteria for the control room design review are being drafted The AIF Control Room
.by the_NRC's Division of Systems Safety.
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Considerations Subcommittee has requested and is planning to meet l
coon with cognizant NRC personnel and their contractor to participate in the formulation of these guidelines and criteria.
7.2 Plant S'afety Status Display Each licensee should be required to define and adequately display in the control room a minimum set of plant parameters (in control
' ~ ~ ~ ~ teTmido-logy',' a state vector) that defines the safety statug,of the nuclear power plant.
The minimum set of plant parameters should be annotated for sensor limits, proces& limits, and sensor status.
The annotated set of plant parameters"should be presented to_.the operator in real time by a reliable, single-failure-proof spstem located in the control room.
The annotated set of plant parameters should also be available in real time in the Onsite Technical Support Center.
The objective of this" recommendation is to require.a, concise set of information that is easily available and assessed by the operator and the shift technical advisor to ascertain the safety status of the' operating process.
The implementation of this recommendation should be undertaken in conjunction with the year-long control room study prmrimisly described, but should be completed by January 1,
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1981, in consonance with the final implementation date for the on-site technical support center recommended in NUREG-0578.
As a further guideline for the development of the safety state vector, the status of the plant process should be designed and instrumented as a function of the various barriers against release of radio-activity.
For example, the two primary barriers are the fuel cladding and the reactor coolant pressure boundary.
Thus, parameters such as primary liquid inventory and coolant radioactivity levels would be principal components of the state vector for these levels of defense.
Similarly, reactor coolant level, containment water level, contain=ent hydrogen content, etc., would be principal components of the state vector for the engineered safety feature levels of defense.
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l, The first sentence is consistent'with a reccmmendation in the AIF Control Room Considerations Subcommittee report.
The Sub-committee'soon will meet with cognizant NRC personnel and con-tractors to participate in developing specific criteria.
7.3 Disturbance Analysis Systems We recommend that the Office of Nuclear Regulatory Research estab-lish a program to evaluate the safety effectiveness of designs of disturbance analysis systems.
This program should consider the evaluation of all pertinent methodologies being used in disturbance o
analysis systems.
The evaluations should be quantitative in nature and include prototype assessments in operating power plant environ-ments.
Experience gained in this program should be used to consider whe-her regulatory requirements should be formulated for the use of disburbance analysis systems in operating plants. _
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[t 5 'p$esent time designs for such'sistems do not exist. '
' Programs have been initiated by botli4PRI and DOE to perform __
Y the' scoping and feasibility study for a plant-wide disturbance analysis system.
Westinghouse and Common' wealth Edison Company have committed in principle, that dependant upon the results from the aforementioned study, a full scale demonstration system be" designed and implemented at Zion operator training simulator and if successful at the Zion, Byron or Braidwood Nuclear Power Plant Stations.
We, therefore, recommend that any program to be performed by the NRC be developed with full recognition of the program already initiated by EP-RI and DOE.
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i 7.4 Manual versus Automatic Ocerations
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We r6co=end that the Office of Nuclear Regu'latory Research formulate a program 3 ".
to establish a technical basis for definitive, licensing criteria for manual and automatic operations for systems, which execute plant safety functions and safety related functions.
The study should include examination of the feasibility of backfit.of its conclusions and recommendations to operating plants.
The
"" ^~ ~C role of-the op.erator should be specifically examined.
Complexity of the safety function, the rapidity of the initiating events, the response time
,.-- available.to diagnose the event and to implement corrective action, and verifica-
- tion of the corrective action should be considered in the program.
TI.e scope of the proposed study includes the operator, the control room, displays and instrumer tation, in addition to the manual and automatic controls that execute safety fur-tions.
The research team should consist of human factors engineers, control ensineers, and nuclear system engin_eers and analysts.
The AIF We agree it. principle with the NRC reccmmendation.
Sub.c.ommitte.e. report provided what we believe are desirable principles for designing a control room.
The firYt principle states,
" Involve the operator as an integral par _t of the process to _
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avoid abnormal conditions or respond properly if they occur".-
We believe that a program to establish criteria for manual and automatic operations'for safety and safety related systems should be joint NRC/ industry effort as the most effective means of saEisfying the NRC recommendation.
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7.5 Standard control Roca Desicn The Instit1Jte' of Electrical and Electronic Engineers (IEEE) has established a
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standards development committee to define design requirements for the standard control. room.
The regulatory staff is represented on the ccomittee.
We recommend that this standards committee expeditiously complete its Work of establishing'st'andard design requirements for. future control rooms.
The
,, ".... _. ~ desigtr requireihent's should consider the lessons learned from the TMI-2 accident as well as the. principles of human-factors engineering for the man-machine
- --' interface.--Upon completion of the standard, the Office of Standards Development
- should evaluate the standard for its acceptability in the licensing pr.ocess, including considerr. tion of its partial applicability to plants under construction.
Until -there -is a - standard. plant, a standard control room is premature.
We believe that a standard related to design methodology would be useful.
The IEEE standards development co:.mittee does not
- h. ave...sufficiently' broad representation from the nuclear industry to define design requirements for a standard control room.
Also,, any
- _s standards developed should be coordinated trith section 7.1 abczy.e.
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REL: ABILITY. ASSESSME'iTS OF FINAL CESIGNS The staff should initiate a systematic assessment of the reliability of safety systems. in operating units and in units in the, late stages of construction using simplified fault and event tree analyses.
Since these assessments go beyond the requirements of current regulations, their completion should not be a condition of licensing for. operation.
The purposes of these assecsments
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would be (1) to audit the implementation of the ' current NRC design requirements by searching for areas that have potential to seriously decrease reliability. -
'and (2) to identify outliers in overall system safety compared with designs previously subjected to this type of review.
Measures to correct any problem
- areas should-be promptly referred to t e cogn zant licensing organization h
i where, in consultation with the Regulatory Requirements Review Committee, backfit decisions are to be promptly reached.
If a particular deficiency is identified and known to exist in several systems or plants, appropriate revisions to NRC design requirements should be made with all licensees and applicants being directed to implement the design revisions in their plants.
Possible approaches would be to assess all systems in One plant or sev.eral systems 1n all plants.
An acceptable combined approach would be to do all systa=s in a few lead plants and then proceed plant; by plant unless particular systems indicated possible. generic problems.
The suspect systems would then by' assessed in all plants', in the manner employed with PWR auxiliary feedwater systems ili' the s'uinmer of 1979.
This recommendaiion would apparently bo satisfied 7
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by the Integrated Reliability Evaluation Pr'ogram currently under development in the Office of Nuclear Regulatory Research with.the previously expressed' concur-
"rence of the Office of Nuclear Reactor RegulatToni-c.c.
Comments:
This recommendation..is to be carried out by the NRC Staff; intend to communicate with them to oitain clarificatUn on the we intentJof the recommendation.
Although the NRC proposes that these assessments be used to evaluate system weaknesses, the NRC 1
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snould also recognize and accept the use of this methodology to evaluate system strengths and permit the removal of nonproductive Reg. Guide and SRP requirements.
The NRC Staff, in implementing this recommendation, should recogni:e the demands already placed on licensee resources and should endeavor to internali e tnis effort within the SRC Staff.
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(7.:cc ennt;en o - contir.ued) iiowever, early interaction with industry experts is essential.
Tne assistahce of our Subccm:aittees on Systems and Equipment Design Criteria and Safety Analysis Criteria is offered.
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'9.
REVIEW OF SAFETY CLASSIFICATIONS AND QUALIFICATICNS
'The owners of operating plants and all plants under construction should be I
requir'ed' to-evaluate the interaction of non-safety and safety grade systems during normal operation, transients, and design basis accidents to assure that cny interaction will not result in exceeding the acceptance criteria for any design basis event.
The review should be systemati and include all non safety d structures under all conditions of normal
" ",_,, components,. equipment, systems, anoperation, antig.ipated operational' occurrenc initiated both within the plant (such as pipe breaks) and from outside the l
---plant (such_as e'arthquakes, other natural phenomena, and offsite hazards).
The intaractions and effects should consider various failure modes including spurious operation,' failure,to operate upon demand, and any unusual or erratic
. operation that might result from exposure to the abnormal process or environ-mental conditions accompanying the event under study.
As a necessary part of this evaluation, proper qualification of safety systems, including mechanical components, s,.hould be verified.
The number of simultaneous failures of non-safety equipment considered should reasonably reflect the expected number of non-safety systems simultaneously exposed during the event under study to conditions 'for. which they were not (fesigned or qualifud..
... : a,. _
Equipment iden'tified as tne potential cause of VTolation of the acceptance criteria for any design basis event'should be appropriately modified to. eliminate
. :or significantly reduce the probability of thtadv_erse interaction. ' Alternatively, the affected' safety systems or structures should b'e modified to cope witPthe intera'ction.
The results of the evaluations should be used to review,*and rodify as appropriate, the plant operating and emergency procedures and operator training.
The Task Force recommends that these studies be completed within a year, at which time licensees should submit proposed schedules for caking the l
riodifications identified in the. evaluations., Completion of this study would l
not be a condition of licensing new plants in the interim of o~ne y~ ear if the f
basis for continued licensing in face of the present unresolved safety issue tn systems interaction is judged by the staff to continue to be valid.
Comments:
1 h'e agree that analyses similar to those requested here could be helpful to promote greater understanding within the industry of the capabilities and limitations of existing equipment to carry out-safety functions.
However, before this recommendation can be implemented, the program described has i
to be further bounded,with priorities established for making I
l stepwise progress.
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they b~ec6me almost limitless.
Results of the recent efforts by Sandia to resolve generic Item A.17 are indicative of this problem.
In or. der to attain an orderly and acceptable resolution
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i of this issue, it is essential that the industry (with periodic 1
communications with the NRC Staff) take a lead role in estab-lishing tne detailed scope of the investigation and in defining
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the acceptance criteria against which the results of the in-vest.igations can be evaluated.
This function will be performed i
- by the CRLS Sub. committee on Systems Interaction.
In-plant implemented,
... f'm'dnt changes or modifications shotTid not be equ /
except in special circumstances of elear safety signi ficanc'e,
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iith this in mind, this recommendation should be a Cate i
gory II item, not Category I as stated.
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CE5IGN FEATURES FOR CCRE-DAM. ACE AND CORE-MELT ACCIDENTS The Task Force rec mmends that the Commission issue within three months a notice of intent to-concuct rulemaking to solicit. comments on the issues and facts relating to the' consideration of design features to mitigate accidents that would result in (a) core melt and (b) severe core damage, but not substantial melting.
Specific areas for comment should include, but not be limited to, the following.
4 r ~-(1)
Are design features. to, mitigate the consequences of either or b'oth of these types"of accidents necessary to provide reasonable assurance that the health and safety of the public are protected?
(2)
In lieu of such features, should additional and supplemental means of l
preventing core damage or core-melt accidents, through improved engineered safety features be required?
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(3) What-should-be the objective of.such design, features? Should the design objectives be a set of specific accep znce criteria (e.g., some limitation on calculated offsite dose) or the reduction of potential offsite exposure that is reasonably achievable?
n (4) bhat should be the, characteristics and functions of such design features?
~ ~ (5) What are-the probabilities and consequences of the various event sequences-that might result in releasing significant amounts of radicactivity to the environment? Which sequences are amena.hle.to interdiction and by
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what means? *
(6) What is the expected effectiveness and performance of' suggested means of reducing the consequences of events in which severe damage or substantial melting of the core occurs, in particular, systems for cont?olled, filtered venting of the containment and for preventing the uncontrolled combustion of hydrogen?
(7) How should other requirements, and in particular those for siting, emergency plans and procedures, training or other related areas, be modified if such design features were required?
_~(8)
What-additional information is required or desirable before setting requirements? What information is available, and what information needs to be developed through experiment, test, analysis, or evaluation?
(9) What should be the final form of the requirement, if any? What should be the implementation schedule for new plants, plants under construction, and cperating plants?
The Task Force reccamends that a proposed rule be published for public ccament within one year of the notico of intent.
Com.aents:
A rule naking for identification of design features for core-dwage and core-melt accidents, as given in the Task Force i
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Ke::rmendation 10, cannot be adequately made until a safety ;oal ass been defined, as detailed in Recommendation 11.
It is not logical to try to iden'tify new design features for core-damage and core-melt accidents without having a safety goal
.2 upon wh.ich~to j~udge the need for additional design features.
With the criteria for the safety goal defined, it may well be that the existing design features coupled with the siting considerations, emergency-action plans, the improved human and plant operational factors, a$d other accident-prevention measures, that are currently being developed,will meet the safety goal without the need to incorporate additional mitigating safety systems.
- Systems and desi-gn features are already in place for
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_These have already been j
c the subj ect of ruler.aking proceedings.
What is needed is careful attention to operating procedures to assure that such systems are al'1 owed to perform their. intended functions.
Because these design features are already in place, it seems relatively useless to conducf further rulemaking on them.
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2 11.
SAFETY GCAL FOR REACTOR REGULATION Th'e 'Cem.tission should undertake with the staff the development and articulation of clear criteria to define the basic safety goal for nuclear power plant regulation.
Since this goal will be used as a benchmark by the staff in defining new regulatory requirements, definitive policy guidance should also be developed regarding the threshold for backfitting of new reqtiirements to existing plants.
The Task Force believes that the goal should be supplemented
" ~ _. _,,-
where possible with quantitative reliacility or. risk criteria, with ifmitations i
TEThqa_tidi on~tneir use to assure Inat sucn criteria do not impede.the
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capability for timely decisionmaking.
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Comments: -
The AIT, through its Committee on Reactor Licensing and of'a Safety, has for many years advocated the establishment
- saf e.t.y..g o a l-f o r r e a c t or r e gu l a t i on.
(gee AIF. Statement on Reactor Licensing, November 1976; AIF CRLS Statement I
on Licensi'ng Reform, June 1977).
Definition of this safety goaF
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sh'ould precede promulgation of additional regulatory re-quirements and would be used for decisions concerning backfit-ting regulato'ry requirements to exist-ing plants.
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The Federal government should conduct a comprehensive risk assessment of all major electric power production programs.
This study would serve to advise the President and Congress of
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the relative risks of mid-term and long-term power alternatives and the relath;iships of these toward achieving a reasoned energy policy implementation.
Thus,. safety goals can be established.
Specification of a safety goal or " acceptable level of risk" l
does not necessarily require the expression of numerical 'value, l
althou;h this should be strived for; rather, for the present, this icans the recognition of the level of safety offered in vi.ts,,e plants des igns, with the Jereig ;ent of scae way in a
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infor-which necessary improvenents in safety derived from new mition~can be identified and incorpora'ted.
In the broadest this means recognition of the comparative risks of
- sense, alternate energy. sources which are used.to produce necessary power.
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~ tee ~ goal should be an interfacing communications statement sufficiently rigorous and technically explicit to be implementable by thh technical community and sufficiently grounded in compara-m tive terms to'be understandable to the public, policy-makers, and other components of the nontechnical ccmcunity.
This also implies a disciplined staff approach aimed at
- issues' -o.f s'afety significance.
Accordingly, although we support the recommendation, we view with con,cern the specific.means of imp 1.ementation suggested, which involves a qualitative goal ibn-sisting of a regime of requirements that are " required for sa.fety".
~
As explained in Section IV of NUREG-05SS, decisions relative to safety would be "
anchored in'ths necessity to'be con-sisIent with and in furtherance of the regulations (which, in turn, constitute an aggregate of requirements which are' required for safety).
This type of qualitative goal appears
~
to be so flexible as to be easily adapatable for any use deemed i
" appropriate" at the moment.
Thus, any new regulatory " improve-ment" would obviously be " required" for safety; and given the need for consistenc.y with'a national goal, individual cases would not r
be relevant.
The " improvement" would become a backfit, uniformly applied, for the sake of consistency rather than for the sake of i
safety.
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To be effec ive, the safety scal,for nu: e:r pcuer cust be anchored in a realistic comparison of available alternatives for
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supplying large quantities of electricity to our nation.'
- Nuclear,
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coal, hydro, gas, oil,' geothermal and other sources should be considered together in terms of pub 1'ic safety.
The resulting
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-safb'ty g o.a l' mus t be based on a comparison of the hazard presented
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by each.of the technologies involved, and to be relevant, it must
~
include a consideration of the public. health and safety impact of wide-]spreadenergyshortages.
Th'e degree to which an energy s'burce is domestic, and the inherent national security that it a
implies must also be' recognized and considered.
A safety goal crea'ted for monitoring nuclear power in isolation, which ignores these factors, will be neither effective nor relevant.
A perspective ~n the comparative risks of electric power o
~
sources is offered in the Ford Founda lott Report entitled p
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Nuclear Power Issues and Choices.
This report expresses several beliefs:
On ba' lance, the risks associated with the current gener,ation of' light water reactors are accept.able; "the local environmental consequences of the nuclear power cyEle in normal operation are not as serious as those from fossil fuel power generation"; and the extremely unlikely upper bound on the possibility of an extremely serious nuclear accident is not itself unacceptable.
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The approach. methods, and organi:atibr. of the 'NRC -staff 'in performing licansing
. reviews of nuclear power plants should be revised to emphasize the. foDcwing objectiiies': -
An overall system level, integrate'8 rehiew that gi'ves full cor.sidera't' ion (1) to operational safety. aspects and provides for a design basis ' accident ass.essment function from event initiation through consequence mitigation, including the rev'iew of emergency. eperating procedures.
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-(.')' Timely.. analysis of. operating experience and. implementation of needed 2-changes derived from operating experience.
in the application of a,, single overal,1 safety goal.
('3) Disci 'i t (4) Continuity of licensing cognizance and responsibility from initial plant Ticensi'ng, thisughout construction and into operation; Technical cversight of Safety Evaluation Reports to assure increased (5) emphasis on safety while still satisfyino the requirements of the -
~
" administrative process of regulation. '
e -(6).Assuran.ce..cf adeqi[dte operations experience and training for the NRC technichi review staff,.especially those stsTf members assigned respon'si.'
bility in accident ' response situations.
(f)' Dedication of adequate resources to the three irincipal functions ofi4hp
~
Office.of Nuclear Reactor Regulation:
reactor licensing, oversight o(
operat'ng reactors, and resolution of generic safety iss~ues.
(8).. Use of' a fdr.yil" procedure for foliowup on questions and requests from,'tne
...... m Advisory Committec on Reactor Safeguards and its individual members.
Comments:
In general, we caution against organizational reform which provides additional layers of reviews or diverts attention away from the principal safety focus.
Item 5, as an example, offers such potential.
Others, such as Item 2, should take cognizance of parallel industry efforts.
We suggest that all of the above objectives be reviewed in terms of their " bureaucratic effect"; for example, Item 8 should assure that significant NRC Staff effort tre not coupled to the. personal preferences of individual ACRS members, but J
rather to the collegial body as a whole.
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Finally, the below objectives should be actively incor-
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pors:ed'into the review process:
Strengthen the authority with the Proj ect Management
.r Staff to arbitrate differences and resolve,witilin NRR, technical and policy considerations related to individual project issues.
Assure a consistent and integrated philosophy
_ among technical review branches and among applica-tion reviews.
Provide impetus to the implementation of value-impact methodology at the beginning of the process of' developing new or revised regTilatory requirements {
instead of at the end of the pYoce.ss.
c Provide impetus to the development and implementa-
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tion of probabilistic risk assessme'nt methods, where applicable, in the regulatory process.
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SCECO:"JIITTE2 ON IF.ER0ENCY RESPONSE PLAN:::::G CO:O:E::TS ON THE FINAL RECO:O*E::DATIC::S OF THE T:C-2 LESSONS LEA?.NED TASK FORCE
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13.
NRR EMERGENCY RESPONSE TEAM The Task Force recommends the establishment of a designated NRR Emergency h
t of emergencies.
The
'Res):ense Team.'.(ERT) to be on immediate call in t e even
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ERT should be.a multi-disciplinary group composed of radiation control, and health'phy' sics.
In the selection of team members, emphasis should be given to applicable operations experience where possible, and the team should be trained and drilled regularly in emergency response.
The Task. force recommends that the Emergency Response Team be identified.and and at least several members of that team be on call by November 15, 1979, relievsd tecp57arily of normal duties to devote full time to the initial liRT task (to be completed by February 1,1930) of identifying resource require-ments, procedures, training, and facilities, including deployment in the Management Team and the Incidence Response Actio The Task Force further recommends that the
'Cos5isTion consider the potential for NRC involve the NRC Incident Response Center.
' ~ ~ ~ ~
in such. response.
+
c.
The Emergency Response Plan developed by the Subcommittee on Emergency Response Planning and distr'ibuted to the industry identifies what the Subcommittee believes should l
October 13, 1979, be the proper NRC role in plant operations during an emergency.
The Subcommittee does not believe that NRC personnel should have During an emer-a direct operational role during an emergency.
gency it is and should be the utility's responsibility to keep the NRC well informed of the operations in progress and planned that might affect the public health and safety.
The proper role of the senior NRC representative at the site should be to act in and support capacity, meeting with the recovery team daily a revicW 0
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or more often as conditions warrant.
If the "RR Emergency i
Response Team and the NRC Emergency Response Center are used O
in i manner to provide support as indicated above then we agree I
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with the recommendation.
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Cuy A. I.rlotto, Pircctor DES, 53 ACTIO!:S f ECESST.RY TO I"rLE*:E::T THE Tit! ACTIO.l PLAN l
S'Jr. JECT:
f:ET.R-TEPJi OPERT.TI"G Lyrf"T.T REr(IIREl:ENTS U
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2 cr. c: mn ry 11, if, cn EE,1 CD, to r.ansic r G.:
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n-~' for ch. nter in HRC's regulations in order to impicr.:cnt the liear-lera i
Operating License Requirements. Tnose participating were:
Unrren !* inners, NRR Jim Stone. IE j
Jim Wolf. ELD A. L. Eiss. SD '
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G. A. Arlotto. SD (Chainnan) 1 Our recomendations are enclosed. Also enclosed is a mcmaranduta from ELD stating "no legal ob.jections to the recomendations" and providing
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- .dditional vierts regerding implem:ntation of these recomendations.
This complctes our assignment.
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Cuy A. Arlotto, Director
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Division of Engineering Standcrds Office of Standards Development bec w/encis: RBMinogue
Enclosures:
GAArlotto
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Actions Necessary to ALJ$sQ Implement the THI Action Minners
!i P1an, Hear-Tcre Operating JStone 9
License Rcquirements JWolf 2.
Memo from GCunningham, ELD, RETYPED FOR MINOR CHANGE -- SEE to GAArlotto, SD dtd 1/25/80 PREVIOUS CONCUP.t2CES I
I DISTRIBUTION SD RF/ ALPHA CENTRAL FILES DES SUBJ/ ALE RF SD TASK NO. N/A OF FICc h._SD: DES /IA____ NRR IE IELD I SD: DES Ams dmd-_ Misinnert___ JRoAe_ _ _ _ @M11 __ _ _ AAArlo_t_to _; _ _ _ _
sunu uce.
--- / 2 3/80 1/23/80 1/ /80 l 1RS/80
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_.1/_2_3/80 h8C Foem 316 42 76) f4RCM 02040 c u. s. cey..,
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D - Rulo Changa Desirable U - Rulo Change Unnec ssary,
Actions Necessary to Implement the TMI Action Plan.
r Near-Term Operatina License Requirements i.
aquirem*nt Applicable Rule Interim CMnge Final Change _
.A.1.1 Shif t Technical Advisor 50.34(b)(6)(1)
Letter to licenisees U, Revise R.G.1.1 50.34(b)(7) and applicants SRP 13.1.1 50.57(a)(3)
.A.1.2 Shif t Supervisor Duties 60.54(1)
Letter U, Revise R.G. 1.8 50.34(b)(6)(1)
.A.1.3 Shif t Manning (1) SRO and R0 in Control Room 50.54(k)
Tech Specs D. Revise 50.54(k) and 50.54(m) 50.54(m)
(2) Administrative Aide to Shift 50.34(b)(6)(1)
Letter U, Revise R.G. 1.8 Supervisor 50.57(a)(3)
(3) Restrictions on Use of Overtime 50.34(b)(6)(1)
Letter U. Revise R.G. 1.8 50.34(b)(7) 50.57(a)(3)
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50.34 b Letter U. Revise R.G. 1.8
..B.1.1 Organization and Management 50.57 ((a))(6)()1)
Criteria (4
[.B.3.1 Safety Eng' ' earing Group 50.34(b)(6)(1)
Letter U, Revise R.G.1.8 or 1.33
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50.57(a)(4) l.B.3.4 Resident Inspector 50.70(a)
U, No change needed I
50.70(b) l.C.1.1 Analysis and Procedure i
m Modificatf ar, O
(l) Small Break LOCA 50.46 Letter U. SRP E
50.34(b)(6)(v) 50.57(a)(3) l
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I Requirement Applicable Rule Interim Change Final Change I.C.1.1 Analysis and Procedure Modification (Continued)
(2) Inadequate Core Cooling 50.34(b)(6)(v)
Letter U. SRP or R.G. 1.33 50.57(a)(3)
I.C.1.2 Shift Relief and Turnover 50.34(b)(6)(iv)
Letter U. Revise R.G. 1.33 l
Procedures 50.57(a)(3)
I.C.1.3 Shift Personnel Responsibilities 50.34(b)(6)(i)
Letter U. Revise R.G. 1.8 50.57 (a)(3)
- I.C.1.4 Control Room Access 50.34(b)(6)(v)
Letter U Revise R.G. 1.33 50.34(b)(6)(iv) 50.57(a)(3)
I.C.2 Vendor Review of Procedures Part 50. Appendix 8 Letter U. None Criteria I & III I.C.3 Pilot Program for Review of 50.70(a)
None U, None 3 elected Er.gency Procedures I.E.1 Licensee Operating Experience 50.34(b)(6)(1)
Letter U. Revise R.G. 1.8. 1.33 Evaluation Capability 50.57(a)(4) 1.70
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I.E.2 Licensee Dissemination of
. Part 55. Appendix A Letter U. Revise R.G. 1.8 Operating Experiences 50.34(b)(6)(iv) 55.10(a)(6) 55.33'(a )(4)
Part 50. Appendix E.
IV.H U. Revise R.G. 1.8 I.G. Training During Low Power Testing 55.10(a)(6)
Letter D. new mle 11.B.1 Degraded Core - Primary System 50.34(b)(2)
Letter Vent Letter D. new rule II.B.2 Degraded Core - Shielding 4
(2).' op GOC 61 D. new nie II.B.3 Degraded Core - Sampling
- 4. b)(2)_ App GDC 61 Letter -
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tequi rement Applicable Rule Int rim Change Final Change I.B.4 Degraded Core -Training 50.34(b)(6)IV Letter D, new rule Part 50 App. E.IV.H
- I.B.8 Degraded Core - Rulemaking 50.34(b)(2)
D Advance Notice of D. New Rul e Part 100 Rulemaking in Process
- I.B.10 Interim Hydrogen Control 50.44 R revision to 50.44 R, revision to 50.44 Requirements for Small Containments 2
[I.C.l.1 - Mini IREP Letter U, None 3
(I.C.I.'J Reliabili.y Assurance Part 50, App. B Letter U, Revise R.G. 1.33
[I.D.l.1 Relief & Safety Valve Test Part 50, App. A Letter U, Revise SRP 3.9.3 GDC 1 (I.D.l.5 Relief & Safety Valve Position Part 50, App. A Letter U, Revise SRP GDC 13 l!.E.1 Auxiliary Feedwater System Part 50 App. A
~ Letter U, Revise SRP 1(1.a.4 GDC 34 Reliability I.E.1.3 Auxiliary Feedwater Initiation
' Part 50, App. A Letter U, Revise SRP 10.4.9 GDC 34 I.E.3 Emergency Power for Decay Heat Part 50, App. A Letter U, Revise SRP 5.4.7 Removal GDC 34 Part 50, App. E. IV.H I.E.4.1 Containment Penetrations Part 50 App. A.
Letter U, already in SRP 6.5.1 GDC 41 I.E.4.3 Containment Isolation Part 50 App. A Letter U Already in SRP GDC 22 I.E.4.5 Containment Purge Part 50. App. A Letter U, already in SRP GDC 54 I.F.2 Inadequate Core Cooling Part 50, App. A Letter U. May require change to Ins trumen ts GDC 13 R.G. 1.97 1.G Emergency Power for Pressurizer Part 50 App. A Letter U, may require R.G. revisior.
GDC 17 i
s 4
Rsquirement Applicable Rule Interim Change Final Change III.A.1.1 Role of NRC III.A.1.5 Communications Part 50. App. E.
Letter U. Revise R.G.1.101 IV.D III.A.2.1 Technical Support Center Part 50. App. E.
Letter U. Revise.R.G.1.101 IV.A Review SRP 6.4. 9.4.1, 9.5.2. 12.2. 12.3. 12.5 for possible revisions III.A.2.2 Onsite Operational Support Part 50. App. E.
Letter U. Revise R.G.1.101 Center IV.A SRP 13.3 III.A.2.3 Near-Site Emergency Operations Part 50. App. E.
Letter U. Revise R.G.1.107 Center IV.A c
III.A.3 Upgrade Licensee Emergency Part 50. App. E Letter U. No change needed Preparedness III.B.3.2 FEMA-NRC Concurrence in State Part 50. App. E.
Letter D. Revise App. E and Local RERP IV.0 also R.G.1.101 III.D.l.3.a Area Radiation Monitors Part 50. App. A Letter U Revise SRP 12.5 (Partial)
GDC 64 f
III.D.2.1 Control Room Habitability Part 50. App. A Letter U. Revise R.G. 1.78 & l.95 GDC 4 III.D.2.2.b Evaluation of Secondary Part"50. App. A Letter U. Revise SRP 9.3.4. 11.3 Side Hazards GDC 61 III.D.2.2.c Improve Auxiliary Building Part 50. App. A Letter U. Revise SRP 9.3.4. 11.3 GDC 61 III.E.1.1 Improve Vent Gas Systems Part 50. App. A.
Letter U Revise SRP 11.3 GOC 60 IJI.E.1.2.a Surveillance Testing Part 50. App. I Letter U. ALARA Tech Specs (Filtration Systems)(Partial)
III.E.2.1.b NRC Monitoring Part 50. App. A.
Letter U
a GDC 64 J
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- l:I ENCLOSURE 2
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UNITED STATES f.
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NUCLEAR REGULATORY COMMISSION y " > v.
WASHINGT ON, D. C. 20555
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MEl10RANDUM FOR: Guy A. Arlotto, Director Division of Engineering Standards Office of Standards Development i
FROM:
Guy H. Cunningham, III Chief Regulations Counsel Office of the Executive Legal Director
SUBJECT:
ACTIONS NECESSARY TO I!!PLEt1ENT TiiE TMI ACTION PLAN NEAR-TERf1 OPERATING LICENSE REQUIREMENTS OELD has no legal objections to the recomnendations of the ad-hoc subgroup e
regarding actionsnecessary to implement the TMI Action Plan near-term onerating I
license requirements. We wish to emphasize, however, certain considerations which should be borne in mind by those who must act on these recommendations.
The informal method of implementation chosen iri the vast majority of instances is a letter to licensees followed by revisions to various regulatory guides.
Because the current regulations are sufficiently broad and general, they provide room for the staff to adopt the interpretation that they may be satis-fied only by meeting the Action Plan's near-term operating licensing require-ments.
Such informal interpretations, however, will not be binding on appli--
cants, licensees, intervenors, Boards, or the Commission, and may be litigated in individual cases.
For example, an applicant would be free to argue te a licensing board that he has demonstrated " technical conpetence" without provid-r ing for a shift technical advisor, since no rule calls for a shift technical l
advisor. Without such a rule, there can be no guarantee that the staff would j
prevail against all such challenges.
Rulemaking, though it has other drawbacks, would eliminate such litigation by imposing binding requirements.
As stated j
sbove, we have no legal objection to the informal methods selected, provided 1
that the risks of repetitive litigation in individual adjudications are under-stood and accepted.
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Where the option of implementation by technical specification changes is selected, pending a rule change (e.g., for SR0 and RO in the control room), there is the additional factor that opportunities for hearing will be created, since l:
technical specification changes are license amendments.
(Of course, these l'
amendments would not have to be pre-noticed if a finding of no significant hazards consideration was made.) Since a rule change is contemplated as the final method of implementation, additional consideration should be given to eliminating the interim stage of amending technical specifications and proceeding directly to an expedited rulemaking.
/
- I H.CunIngh3, I
gg$.hqppi Chief Regu ations Counsel Office of the Executive Legal Director
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g NUCLEAR REGULATORY COMMISSION
.e WASHINGTON, D. C. 20555 o, Yi
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..o March 5, 1980 MEMORANDUM FOR:
Ed Hant han FROM:
Al K s.
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SUBJECT:
AIF WORK NG GROUP EVALUATION OF TMI ACTION PLAN PRIORITIES Cookie, Dennis and I have taken a look at the AIF Working Group (WG) evaluation.
We summarize below our initial reactions.
I understand the staff Steering Group (SG) will be reflecting a more complete analysis of the WG evaluation in draf t number 3 of the Action Plan.
In general terms, the WG approach to priorities seems no better (or worse) then that used by the SG, although the results for some items are sub-stantially different:
the WG estimated costs to industry substantially greater and benefits generally less than those estimated by the SG.
The WG ranking does provide some additional insights, including unit and total costs and, in some cases, distributions of success probability and implementation times, which may be of help in the final assessment of which proposed actions should be deleted and what schedules should be adopt,ed for those approved.
-- Almost half the items fell in the third priority, a category of items the WG believes should be dropped from the Action Plan.
The most interesting aspect of the report are the WG cvaluations of several items which in whole or in part, are considered by the WG to be contrary to safety or, in some circumstances, to have that potential:
. Requiring B.S. degrees for operators would lead to excessive turnover.
Plant maneuvers for drill purposes would present unnecessary challenges.
Rapid changes in organizational requirements could confuse and compli-cate organization effectiveness.
With direct data links, offsite government personnel might tend to direct or influence plant response action without sufficient knowledge.
l Increased shif t manning might overload training staff and reduce training l
in other areas.
Automatic status monitoring may engender a false sense of security and I
decreased alertness by the operators.
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Ed Hanrahan.
Control room design changes, if introduced too rapidly, may increase chance of operator error.
A plant safety parameter display console may be relied on excessively and operator proficiency with the operating panels may be reduced.
Automatic RCP trip might be spuriously c :tivated.
Automatic closure of the PORV block valve may be detrimental in certain accident sequences or even initiate new sequences.
l A couple of references to WASH-1400's bottom line numbers on reactor risks suggests a frame of reference by the WG which the Commission rejected a year ago.
cc:
Dennis Rathbun Cookie Ong e
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6