ML19254E121
| ML19254E121 | |
| Person / Time | |
|---|---|
| Site: | 05000471 |
| Issue date: | 09/20/1979 |
| From: | Beverly Smith NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Burt L MASSACHUSETTS, COMMONWEALTH OF |
| References | |
| NUDOCS 7910310120 | |
| Download: ML19254E121 (1) | |
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NUCLEAR REGULATORY COMMISSION
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September 20, 1979
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' C Laurie Burt, Esq.
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Environmental Protection Division One Ashburton Place,19th Floor Jf Boston, Massachusetts 02108 In the Matter of Boston Edison Company, et al.
(Pilgrim Nuclear Generating Stition, Unit 2)
Docket No. 50-471
Dear Ms. Burt:
By letter of September 4,1979 you requested a copy of the staff document which prompted the Applicants' submittals of August 14 and 22,1979 to the staff, and a copy of an August 20, 1979 memorandum from H. R. Denton to the NRC Comissioners concerning Resumption of Licensing Reviews for Nuclear Power Plants.
According to Mr. Licitra, Project Manager for Pilgrim Unit 2, the Applicants are responding to the staff's oral request for clarification and augmentation of information previously submitted concerning the Applicants' commitment to TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations (NUREG-0578).
The August 22, 1979 submittal also contains responses to the above-described memorandum from Mr. Denton.
Per your request, I have enclosed a copy of Mr. Denton's memorandum.
Sincerely,
<a-Barry H. Smith Counsel for NRC Staff
Enclosure:
As stated cc w/ encl:
See Pilgrim Service List 1235 189 7910310k N
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.,o ( j NUCLEAR REGULATORY COMMISSION W ASHINGTON. D. C. 20555
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August 20, 1979 itEMORANDUM FOR:. Chaiman Hendrie Comissioner Gilinsky Comissioner Kennedy Comissioner Bradford Comissioner Ahearne 4
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Lee V. Gossick Executive Director for Operations FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation SUBJEC.T:
RESUMPTION OF LICENSING REVIEWS FCR NUCLEAR PCWER PLANTS In May of this year I described a realignment of current and near-tem priority tasks within the Office of Nuclear Reactor Regulation (NRR) to deal with activities relating to the accident at Three Mile Island (see SECY-79-344).
One consequence of the realignment was a temporary delay in the processing of operating license and construction pennit applications for nuclear plants pending completion of certain TMI-2 related tasks.
The, short-tem TMI-2 tasks are essentially complete, as summarized below, and based on the results of these efforts I have decided to resume steff licensing act,ivities on pending construction pemit and operating license applications.
It is my judgment that the TMI-2 related actions being taken by NRR on licensee emergency preparedness (see SECY-79-450), operator licensing (see SECY-79-33-E), bulletins and orders followup (primarily in the areas of auxiliary feedwater system reliability; loss of feedwater and small break loss-of-coolant accident analysis; emergency operating guidelines and procedures; and operator training), and short-tem Lessons Learned, if accomplished generally on the schedule we have selected, are necessary and sufficient for the continued safe operation of operating plants and for the resumption of staff licensing activities on pending construction pemit and operating license applications.
It is my intent to bring the staff's first completed review of a pending operating license application to the Comission for review prior to staff issuance of the license. The Lessons Learned Task Force and I also have considered whetner the actions associated with these activities'would foreclose other actions that subsequently may be shown to be necessary by the Lessons Learned Task Force, tne President's Comission or the NRC Special Inquiry.
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We have no indication that they will.
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The Commission The principal element of the composite of staff activities listed above is the completion of my review and the ACRS review of the first report of the TMI-2 Lessons. Learned Task Force (MUREG-0578). The Task Force report contains a set of recommendations to be implemented in two stages over the next 16 months on operating plants, plants under cor.struction, and pending construction permit applications.
The Task Force recommended 20 licensing requirements and three rulemaking matters in 12 broad areas (nine in the area of design and analysis and three in the area of operations).
All but one of the 23 recommendations had a majority concurrence by the Task Force.
The Task Force concluded that implementing its recommendations would provide substantial, additional protection which is required for the public health and safety.
The Advisory Comnittee on Reactor Safeguards has completed its review of the Tas.k Force report. The several public meetings of the ACRS subccmmittee on TMI-2 and the public meeting of the full comnittee on August 9 provided an opportunity for the presentatien and discussion of public comments on the report. The ACRS letter of August 13, 1979, to Chairman Hendrie states that the Committee agrees with the intent and substance of all the Task Force reccmmendations, except four upon which the Comnittee offered constructive comments to achieve the same objectives articulated by the Task Force. The Committee also noted that effective implementation will require a more flexible, perhaps extended, schedule than proposed by the Task Force. A copy of the ACRS letter is provided as Enclosure 1.
ThekCRScommentsonNUREG-0578concentrateonfouroftheTaskForce recommendations. These are: (a) the revision of limiting conditions of operation to require plant shutdown for certain human or procedural errors; (b) the inerting of MKI and II BWR containments; (c) the provision of recombiner capability at operating plants tnat do not already have it; and (d) the addition of a shif t technical advisor at each operating plant.
The first three of these matt'rs require Commission rulemaking, and it is a straightforward matter for the staff consider the comments in the process of developing the required Commission papers.
I will assure that is done.
It is my intent to ask the Office of Standards Development (50) to proceed expeditiously with a Commission paper proposing a new rule on limiting conditions of operation (item a, above).
I will ask SD to include in the paper the alternative approach recommended by the ACRS, and one other approach that I think merits consideration. My alternative would amend the Task Force recommendation so as to differentiate between an isolated occurrence and a repetitive pattern. For example, the forced shutdown aspect of the Task Force recommendation could be raserved for a repeat violation within a relatively short time period, such as two years.
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The Commission,
In the case of the two hydrogen control matters (items b and c, above), I intend to follow the' advice of the ACRS by asking SD to delay completion of the required staff papers for proposed rulemaking until after receipt and review of the final report of the Lessons Learned Task Force, now scheduled for completion in mid-September.
It is likely that the inerting and recombiner requirements recommended by the Task Force will be included in the eventual solution to the hydrogen control problems encountered in the TMI-2 accident. However, in view of the short time until the availability of the overall hydrogen control recommendations by the Task Force, I agree with the ACRS that it is best to not dilute staff effort in this area by prcmpt pursuit of the two short-term recommendations, one of which was a minority view of the Task Force for these same reasons.
The ACRS comments on the shift technical advisor (item d, above) have resulted in our reassessment of the possible means of achieving the two functions which the Task Force intended to provide by this requirement.
The two functions are accident assessment and operating experience assessment by people onsite with engineering competence and certain other characteristics.
I agree with the Task Force that the shift technical advisor concept is the preferable short-term method of supplying these functions. However, I have concluded that some flexibility in implementation may yield the desired results if there is management innovation by individual licensees. The Task Force has prepared a statement of functional cnaracteristics for the shift technical advisor that will be used by the staff in the review of any alterpatives proposed by licensees.
It is previded here as Enclosure 2.
In addition toicemmenting on four of the Task Force reccmmendations, the ACRS letter of August 13 reccmmends three additional instrumentation requirements for short-term action. These are containment pressure, containment water level and containment hydrogen monitors designed to folicw the course of an accident.
I agree with these recommendations. The Task Force has prepared descriptions of these recuirements in the same format as Appendix A of NUREG-0578. They are provided here in Enclosure 3.
I have also decided on ene further licensing requirement for short-term action.
It is a requirement for remotely operable high point venting of gas fr9m the reactor coolant system. The Task Force has prepared a description of this requirement; it is provided here in Enclosure 4.
The Task Force had previously deferred this item for further study, but it is my judgment that design efforts by licensees can and should be initiated ncw.
Finally, the Task Force has comoiled a set of errata and clarifying comments for NUREG-0578.
It is proyided here as Enclosure 5.
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The Comission In summary, the Task Force recommended promot licensing action on 20 items (excluding the three rulemaking matters).
I have added the three additional requirements recommended by the ACRS in its August 13 letter and one more on the basis of my own review. This Office will issue letters to all ooeratina plant licensees and all constructicn pennit and ocerating license applicants within the next two weeks requiring them to commit within 30 days to meet the total of 24 licensing requirements on the implementation schedule provided here in Enclosure 6.
Another letter to be 'ssued at approximately the same time, will state the requirements fic,ing from the work by the Bulletins and Orders Task Force on operating plants which also need to be picked up on the license applications.
Several licensees have advised that some of t!.e hardware changes required in NUREG-0578 can be acccmplished at much lower cost during springtime refueling outages in 1980.
For good cause shown, we intend to consider such flexibility in the implementation schedules. The end date for full implementation of all licensing requirements has not been changed from the January 1,1981, date recomended by the Task Force. The implementation dates for the Comission rulemaking actions will be established in the course of rulemaking.
hkh Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
- 2. Alternatives to Shift Technical Advisors
- 3. Instrumentation to Monitor Containment Conditions
- 4. Installation of Remotely Ooerated High Point Vents in the Reactor Coolant System
- 5. NUREG-1578 Errata
- 6. Implementation of Requirements for Operating Plants and Plants in OL Review cc: Mitchell Rogovin Saul Levine Robert Minogue Victor Stello William Dircks Carlton Kamerer ACRS 1235 193
ENCLOSURE 1
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s was>uscTon. o. c. 20sss g% f August 13, 1979 Honorable Joseph;M. 'Hendrie Chairman U.S. Nuclear Regulatory Ccemission Wdshington, D.C.
20556 SHCRT-Tere PfECrd?INCATICNS CF T4I-2 LESSCNS EZARNED TASK FCRCE
SUBJECT:
Dear Dr. Hendrie:
9-11, 1979, the, Advisory Ccemittee on During its 232nd meeting, August Reactor Safeguards ccepleted a review of the short-term recccmendations of the T11-2 Lessons I4arned Task Force as reported in MREG-0573.
These recommendatic~ had been reviewed, in part, by an ACRS Subcommittee at a meeting in Wasnington, D.C., on July 27, 1979. During its review the Committee had the benefit of discussions with menbers of the Task Force.
Comments frem representatives of the nuclear industry were also considered.
In its review, the Committee has noted that the reccomendations in NuuG-0578 are those deeced by the Task Force to be required in the short term to pro ride substantial additional protection for the public health and safety.
The Committee has considered both the recccmendations themselves and the schdules proposed for their implementation. Fegarding the latter, the Ccenittee believes that the orderly and effective implenentation and the appropriate le' vel of review and approval by the hRC Staff'will require a somewhat more flexible, and in some cases more extended, schedule than is implied by NLREG-05781 With regard to the requirenents themselves, the Ccemittee agrees with the intent and substance of all except those discussed below.
2.1.5 Post-Accident Hvdrecen-Centrol Systems The Committee agrees with the reccomendations relating to dedicated a_.
penetrations for external recccbiners or purge systems for operating plants that have such systems.
- b. and c.
2.e majority of the Task Force has recccmended rule-making to require inerting of EWR.vark I and II reactors. A minority of the Task Force Pas recer.vrended rule-making to require that all cperating 1.'.ght water reactors provide the capability to use a hydrogen recccbiner.
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Honorable Joseph M. Hendrie August 13, 1979 The Committee believes that qJestions relating to hydrogen generation during and following an accident, the rate and amount of generation, the need to control it,.and the means of doing so, need to be reexamined. The Task Force has advised the Committee that it is considerity this question further in connection with its lorger-term recccmendations which are sched-uled to be ecc.plsted by September,1979. The ACRS believes that decisions concerning pssible additional measures to deal with hydrogen should be deferred pending early evnluation of the fortheccing lcrger-tern Task Force reccemendations.
2.1.8 Instrumentation to Follcw the Course of an Accident With regard to instrumentation to follcw the course of an accident, the ACRS believes that contairment pressure, containment water level, and on-line menitoring of hydregen concentration in the centairment 1ould also be considered for implementation for all eperating reactors on the. same schedule as that recccmerded by the Lessons Learned Task Force.
2.2.1.b Shift Technical Advisor to Committee agrees ccepletel" with the two clcsely related objectives of this recccmendation. Che rela a to the presence in the control room dur-ing off-normal events of an individual having technical and analytical capability and dedicated to concern for safety of the plant. 2e other relates to the need for an on-site, and perhaps dedicated, engineering staff to review and evaluate safety-related aspects of plant design and operation.
. We achievement of these cbjectives will contribute significantly to the safe operation of a plant.
We Committee believes that there may be difficulty in finding a suf ficient number of people with the required qualifications and interest in shift work to fill the Technical Advisor psitions. We Ccemittee therefore believes the solution proposed by the Staff should not be mandatory but that alternate solutions also should be considered.
2.2.3 Revised lim dnc Conditiens for Ooeration te Ccemittee agrees with the findings of the Task Force that there are too many htnan. or operational errors resulting in the defeat of an entire safety system, that the number of such occurrences should be and can be reduced, and that the ultimate responsibility for doing this must rest with the licensee.
The Committee, however, is not convinced that the Task Force proposal is the best or only,cy to increase the licensee's av. reness of the 1235 195
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t Honorable Joseph M. Hendrie August 13, 1973 need to improve operational reliability, and stqgests that measures short of shutdem, stx:h as a rule that requires actions similar to those of a show-cause order, may be equally effective.
Sincerely, A
Max W. Carbon Gairman
References:
1.
NUREU-0578, "24I-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," Of fice of Nuclear Reactor Fegulation, U.S.
Nuclear Regulatory Commiss'on, July 1979.
2.
Letter, D. Knuth, Presidenc, KMC, Inc., to Harold Centon, Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Ccemis-sion, August 8,1979, subject: T4I-2 Lessons Learned Task Force Report (NUREG-0578).
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3.
I4tter, Stanley Ragone, President, Virginia Electric and Powr Company, to Joseph M. Hendrie, Gairman, U.S. Nuclear Regulatory Ccmmission, August 8,1979,
Subject:
Iassons Learned Task Force on T4I-2, NUREG-0573.
4.
Letter, Floyd W. Lewis, Qairman, M Hoc Nuclear Oversight Committee, to Harold R. Denton, Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Ccemiscsion, August 1,1979,
Subject:
Iassons Learned from T4I-2.
5.
Latter, American Nuclear Society, ANS-3 Committee, to Joseph M. Hendrie, Gairman, U.S. Nuclear Regulatory Commission, August 2,1979,
Subject:
Iassons Learned Task Force Status Repart NUREn-0578.
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UNITED STATES NUCLEAR REGULATCRY CCMMISSION ADVISORY cCMMITTEE ON REACTOR SAFEGUARCS
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......'3-August 15, 1979 MEMORANCUM FOR:
Chaiman Hendrie FRCM:
Raymord F. Frale
_xecutive Director, ACRS
SUBJECT:
ADDITIONAL REFERENCES TO ACRS LETTER ON SHORT-TERM RECCKMENCATICNS OF TMI-2 LESSCNS LE.RNED TASK FORCE DATED AUGUST 13, 1979 The attached revised Page 3 of the subject letter should be substituted for the one which was originally sent to you.
This page incorporates additional references 6, 7, and 8.
Attachment:
Revised Page 3 cc:
Comissioner Gilinsky C0missioner Kennedy Comissioner 3radf:rd Cemissioner Ahearne
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Honorable Joseph M. Hendrie August 13, 1979 need to 1.: prove operational reliability, and suggests that measures dort of shutdom, su:h as a rule that recuires actions similar to th:se of a show-cause order, may be equally effective.
Sincerely,
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Max w. Carbon Cairman
References:
1.
NmIn-0 573, ""'4!-2 Lessons Learned Task Force Status Report and Short-Term Reccccendations," Cf fice of Nuclear Reactor Regulation, U.S.
Nuclear Regulatorf Commission, July 1979.
Letter, D. Fauth, President, IC4C, Inc., to Harold Centon, Director, 2.
Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Cocais-sien, August 3,1979,
Subject:
"MI-2 Lessons Learned Task Force Report (NGEn-C573).
3.
I.atter, Stanley Ragone, President, Virginia Electric and Power Company, Commission, to Joseph M. Hendrie, Qairman, U.S.. Nuclear Regulatorf August 3, 1979,
Subject:
I.assons Learned Task Force on 'IMI-2, NURC-C578.
4.
I.atter, Floyd W. Lewis, Gairman, Ad Hec Nuclear Oversight Coenittee, to Harold R. Centon, Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory C:caission, August 1,1979,
Subject:
Iassons IAarned frec 74!-2.
5.
Iatter, American Nuclear Scciety, ANS-3 Coccittee, to Joseph M. Hendrie, Cairman, 'J.S. Nuclear Regulator / Cocaission, August 2,1979,
Subject:
Lessons Learned Task Force Status Repsrt SURIC-0579.
6.
Letter, Robert S:alay, Atomic Industrial For.mt, Inc. (AIF), to Harold Centon, Director, Of fice of Nuclear Reactor Regulation, U.S. Nuclear Regulatorf Commission, August 2,1979,
Subject:
"':MI-2 Lessons Learned Task Force Status Report and Short-Te:m Reccreendations," (NURIn-0578),
7.
Pepert by the AIF Policy Coenittee on Follow-up to the ':hree Mile Island Accident, July 5,1979.
8.
Memorandten, C. G. Long, Lessons Learned Task Force Member, to R. J. Mattson, Directo r, 3.I-2 ' assons : earned Task Force, July 30, 1979,
Subject:
Feview of' LF.Es for Loss of Safety. :nction Cue to Personnel Irror and Cefective Procedures, (50-320).
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ENCLOSURE 2 ALTERNATIVES TO SHIFT TECHNICAL ADVISORS The recommendation by the Lessons Learned Task Force that an on-shift Technical Advisor be required at operating nuclear power plants has received much comment and attention by the ACRS and industry representatives since NUREG-0578 was published. Several alternative approaches have been suggested.
The ACRS has advised and the Director of NRR has decided that alternatives be considered and approved if found by the staff to satisfactorily acccmplish the functions described by the Task Force for the Shift Technical Advisor. Ai an aid to evaluating alternatives, a more comprehensive discussion of the purpose and basis of the Task Force recommendation is provided below. The discussion is in terms of the two principal functions intended to be accomplished and the characteristics thought to be necessary to effectively accomplish these functions.
,It is intended that the licensing review staff make use of this discussion in eva10ating alternatives proposed by licensees and license applicants.
Introduction As stated in NUREG-0578, the Lessons Learned Task Fcrce has concluded that the need for improved operaticns is the most important lesson learned from the accident at TMI-2. Cne key element so far identified is the need to improve the capability in the control room to recognize and diagnose unusual events.
Over the next several years, impruvements in the capability of the reactor operations staff to respond to unusual events can and will be sought through improvements in plant desiga, operating precedures and the qualification and training of operators.
Improvements in plant design are expected to include imcrovemer.ts in the area of human f actors, esoecially incrovements in display 1235 199
2 and diagnostic systems available to aid operators. For example, the Task Force made a short term recommendation for improvement of the means of assessing inadequate core cooling. The Task Force also made short term recccrendations for improvements in emergency procedures and preparations by the plant operations organizaticn. The purpose of these recorrendations is to assure that the operators and the onsite operational and technical support personnel are organized both administratively and physically in an effective manner.
In addition, improvements in the licensing requirements for operators have been recocaenced to the Ccmission. Over the ccming months, it is likely that further increases in qualification and training requirements for operators will be developed by the industry's recently anncanced fluclear Operations institute for implementation over the next several years. Because these changes are necessary but difficult to achieve rapidly, the Lessons Learned Task Force has recommended
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the use of Shift Technical Advisors as a method of immediately improving the operating staff capabilities for response to off normal conditions and for evaluating operating experience.
The consensus of the Task Force is that there are two necessary improvements in the capability to assess the status of a plant during~ unusual conditions such as a transient or an accident, to realize the significance of the available information such as instrument readings, and to take appropriate action. First, there shocid be an accident assessment capability based on a comprehensive education in engin-eering and science subjects related to nuclear power plant design and on training and experience in the dynamic response of the specific plant. This capability I
must be rapidly available in the control recm in the event of an accident. Second, there should be a capability to maintain and upgrade safe plant cperatiens through the cognizance and evaluation of applicable cperating experience by an engineering group with diverse technical kncwledge, ex::erience, and pers::ective in relevant areas such as electrical, mechanical and 1235 200
3-fluid systems and human factors. The addition of Shift Technical Advisors to the olant operating staff is an acceptable means of supplying both of these functions. Alternative manning and crganizational schemes will be considered and will be evaluated for satisfaction of the qualifications, training and duty assignment criteria discussed below.
Discussion In developing the recommendation for the Shift Technical Advisor, the Task Force concentrated on the two functions that needed to be provided, namely, an accident assessment function and an operating experience assessment function. The proper performance of these functions requires the provision of certain characteristics described in the folicwing paragraphs.
A.
Accident Assessment Function 1.
General Technical Education The technical education of at least one person in the control rocm under
'off normal conditions should include basic subjects in engineering and science.
The pdrpose of this education is to aid the operator in assessing unusual si*uations not explicitly covered in the current operator. training.
The following is a tentative list of areas of kncwledge that are considered to be desirable:
Mathematics, including elementary calculus Reactor physics, chemistry and materials Reactor thermodynamics, fluid mechanics, and heat transfer Electrical engineering, including reactor control theory These areas of knowledge should be taught at the college level and would be equivalent cc about 60 semester hours. Although a college graduate engineer would have many of these subjects and more that would not be essential, some engineers might be deficient in a few of these specific areas, e.g., reactor 1235 201 e
-4 physics. Although the time to teach these subjects to a licensed senior reactor operator could be as short as two years, depending on the scope and content of the subjects, the selection of a graduate engineer would likely be a more rapid means of fulfilling this characteristic.
2.
Reactor Operations Training All persons assigned to duties in the control room should be trained in the details of the design, function, arrangement and operation of the plant systems. This training is necessary to assure that the meaning and significance of instrument readings and the effect of control actions are known. A licensed operator or supervisor of an operator would not be required to have further training in order to fulfill this characteristic. A graduate engineer not previously licensed or trained as an operator or senior operator would require additional training in order to fulfill this characteristic.
3.
Transient and Accident Response Training In addition to the training in normal operations, anticipated transients, and accidents presently recuired cf operators and senior operators, one person in the control room under off normal conditions shculd be trained to recognize and react to a wide range of unusual situations including multiple aquipment failures and operator errors. This training should not be limited to written procedures or specific accident scenarios, but should include the recognition of symptoms of accident conditiens such as complex transient resconses or inadequate core cooling and possible corrective acticnn.
The purpose of this training is to broaden the ability for prompt rc agnition of and response to unusual events, not to modify the instinctive, rapid procedural response to transients and accidents provided by reactor operators. The training is required in recognition of the fact that real accidents inherently are initiated and acc=canied by unusual ano unex:ectec events. The training is also to em:hasize 1235 202
-S-need to fccus on the essential parameters that indicate the status of the core and the primary coolant boundary. This additional training would take up to a year to ace:mplish for a person not already exrerienced in nuclear plant transient and accident analysis or evaluation. Both inexperienced graduate engineers and currently licensed operators would require additional training to fulfill this characteristic.
4.
Detachment from Operations The plant response assessment function requires a measure of detachment frcm the manipulation of controls or imediate supervision of operators. This is intended to provide the perspective and the time for assessing plant conditions and advising on appropriate cperator actions.
It has been called a safety monitor characteristic. Currently only three operators would normally be in the control recm at the time an unusual event occurred, and it is allowed that at times there would be fewer. This number is only enough to satisfy the demands for prompt control and supervisory actions under off normal conditions. The time necessary to make a considered assessment and permit independent monitoring of plant safety require one more person in the form of the Shift Technical Advisor or soma alternative in the control rocm.
5.
Independence from Operations In order to provide both perspective in assessment of plant conditiens and dedication to the safety of the plant, this function should have a clear measure of independence from duties associated with the commercial operation of
'he plant.
In an accident situation where ccamand authority should not be diluted, complete independence is not desirable and is not necessary to the safety assessment function.
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- 6.
Availability This capabilii.y should be readily available in the control room, preferably immediately at all times, but at most within ten minutes. Having this capability on duty for each shift is the best approach.
B.
Operating Experience Assessment Function 1.
Independence from Operations A measure of independence is required to provide for e'ffective safety monitoring of operating experience at the individual plant and at plants of
! ke design. The assessment of operating experience at the assigned plant and other similar plants and the routine monitoring of the safety of plant operations is usually compatible with and necessary for efficient operations. However, the demands of ccmmercial operation can sometimes distract from or appear to override safety judgments. An independent monitoring of the safety of plant operations is intended to counter-balance the immediate and pressing needs of commercial operation.
2..
Ded1 cation Personne] should be dedicated to the function of safety monitoring of operating experience as their primary responsibility and duty. Although reactor operating personnel have a comitment to safety that derives frca self interest as well as regulatory requirements, it is only cae of two primary responsibilities,
the other being the continucus production of power. The assignment of safety evaluation of operating excerience as a primary responsibility for certain specified individuals will reduce potential conflicts and assure adequate time to discharge the duties.
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. 3.
Diversity of Technical Knowledge The technical knowledge of those assessing operating experience should be diverse and encompass all technical areas important to safety. The types of problems that can affect safety include all areas related to the design and operation of nuclear power plants; e.g., mechanical, electrical and fluid systems and reactor physics, chemistry and metallurgy. Recognition and under-standing of a problem and its significance requires some knowledge in the relevant technical specialities and cannot depend solely on the descriptions and judge-ments of the persons identifying and reporting the problem. Because of the broad scope of possible technical areas and the possible interactions of components, equipment and systems, the people engaged in operating experience review should have experience in areas usually designated as systems engineering.
They should also be graduate engineers, or equivalent.
In addition, because of the importance of operator actions in the safety of plant operations, familiarity with or routine access to persons with the principles of human engine 3 ring or human, factors should be provided.
Alternatives As discussed in NUREG-0573, several alternative means of providing the accident assessment function were considered by the Lessons Learned Task Fcrce. They were:
1.
Upgrade the requirements for reactor operators and senior reactor operators tc include more engineering and plant respcnse training.
2.
Provide additional on-shift personnel with science or engineering training and specific traning in plant design and resocnse.
3.
Provide on-call assistanc-to the control room by identified personnel in the plant engineering organization having the training described in alternative 2.
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8-Althougn the Task Force initially assumed that the accident assessment function would be combined with the operating experience assessment function, it is possible that the two functions could be separated. Scme have suggested that people with the education, training, and experience required for both the operating experience assessment function and the safety monitoring function would be more easily obtained and retained if not required to work on shift.
Others believe that such people can be retained if sufficient incentives are provided. The advantages and disadvantages of these alternatives are discussed below. Although no alternative other than a group of dedicated Shift Technical Advisors has so far been found acceptable, it is possible that innovative improve-ments in.the other alternatives could be found acceptable.
Discussion of Alternatives 1.
Upgrade the training and cualifications of the senior reactor coerator.
This alternative would require no change in the present number or organization of control room ;erators. The debilitating feature ~of this alternative is that the senior' operator would be busy directing the reactor operators or taking actions himself during an accident and not have sufficient time or perspective to make the desired assessment of plant conditions; i.e., perform the safety monitor function. This arrangement would also not provide a clear independence from cormiercial operation. However, the capability would be readily available when needed.
It is unrealistic to expect the senior operator to fulfill the operating experience assessment function. A separate group could be established to accomplish that functicn on the day shift when interaction witt. offsite experts and utility management would be enhanced.
If schemes are proposed to accomplish the two functions separately, then they should include mechanisms 1235 206
~
9 for sufficient coupling of the two to assure continuous feedback of and ready access to the knowledge being acquired in operating experience evaluation.
2.
Additional on-shift personrel This alternative wculd require the addition of one person to the on-shift control rocm staff.
If the person is to be a Shif t Technical Advisor, no license would be required, thus making the position easier tc fill quickly. However, detachment from first-line comercial operations decisions can be attained by either a line or advisory position. For example, instead of the Shift Technical Advisor proposed by the Task Force, there may be acceptable methods of using a Shift Engineer, who normally has authority over a Shift Supervisor, to perform the accident ssessment function.
Either approach would utilize people on shift so they would be readily available. Since the Shift Engineer would have normal duties other than operating experience assessment, a separate day shift group would be required to fulfill that function if the shift engineer was found to be an acceptable source of the accident assessment (safety monitor) function.
3.
On call assistance This alternative would require no sdditional on-shift personnel. Others have susggested that provision of the recomended technical education and training would be most easily acccmplished with this alternative since degreed engineers with intimate knowledge of the plant design basis and accident response character-istics are available in the utility technical staff. Since these personnel would be remote frcm the control room, a requirement to be licensed dces not appear to be consistent. Knowledge of accident response might also be more easily found among vendor personnel who have extensive experience in accident analysis and systems design. This alternative also provides detachment from actual operation and scme independence from cwercial cperation. Pcwever, these pecole aculd 1235 207
not be readily available when needed. The use of utility or vendre oersonnel not.at the site would increase the difficulties of communication.
.t though there is need for backup assistance from these other organizations, it is doubtful that they would be able to provide for the prcmpt response needs of the accident assessment functicn and they do not have sufficient plant unique experience and f amiliarity to satisfy the operating experience assessment function.
1235 208
InstrumenM,on to Monitor Containment Conditions During tb t Course of an Accident 1.
INTRODUCTION General Design Criterion 13, " Instrumentation ano '
r 1," of Appendix A to 10 CFR 50, requires instrumentation to mon tariables "for accident conditions... including containment and assaciated systems."
Specific requir ints are included in Standard Review Plan Section 6.2.5,
" Combustible Gas Control in Containment," for the capability to monitor hydrogen concentration in the containment atmosphere.
Instrumentation to sense or monitor containment conditions already exists to some degree (e.g., automatic containment isolation on high containment pressure at TMI-2). However. it is clear that all information necessary to assess the response of the containment to the accident conditions at TMI-2 was not available to :.he operator.
It has bebn the contention of some applicants that General Design Criterion 13 applies to only those accidents listed in Chapter 15 of Regulatory Guide 1.70.
Again, based on conditions experienced at Three Mile Island, it is clear that situations can arise which produce contair. ment conditions beyond those postulated for the Chapter 15 events.
1235 209
_ p,_
2.
DISCUSSION Approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the start of the accident at TMI-2, a 28-psig pressure spike occurred in the containment building. Although it is now believed that the pressure spike was due to the rapid burning of hydrogen gas in the containment atmosphere, the staff on duty in the control room apparently did not attach any special significance to the pressure spike at the time. At the time of the occurrence, the plant staff attributed the event to various causes, including electrical problems and relief valve opening.
It is now known that the pressure spike represented a much more serious condition within containment and the pressure indication itself could have been, but was not then accepted as, critical information to the olant operators. The events at Three Mile Island clearly reaffirm the need for containment pressure indication in the control room.
Furthermore, it is clearly cost effective and necessary that the instrumentation range inci'ude the expected failure level for the containment.
1 The sequence of events during the accident at Three Mile Island indicate a second item of information which could have been, but was not immediately accepted as, critical information in the diagnosis of the accident.
This information was the free liquid inventory in the containment building.
During the accident, reactor coolant drain tank quench water and primary coolant water vented through the drain tank relief valve and flowed to the 1235 2i0
. reactor building sump. Water within the containment sump was then discharged to the auxiliary building sump tank and thus resulted in scme transfer of radioactive material outside of the containment building.
Because sump pump operation was expected several times a day before the accident due to routine accumulation, the transfer process was not recognized as an indication of contaminated water in containment.
Furthermore, the accumulation of water in the TMI-2 containment probably contributed to equipment failure due to flooding.
The events clearly establish a need for accurate containment water level indication in the control room, with instrument ranges which include accident flooding levels.
The third item of information which was subsequently considered to be of critical importance in determining containment conditions at TMI-2 was the hydrogen concentration in the containment atmosphere. The hydrogen gas was produced as a result of the reaction of zirconium metal and primary coolant water ip the reactor core. The gas was vented, to some extent, from the reactor coolant system to the containment atmosphere. The free hydrogen in containment further resulted in a rapid burn and pressure spike event in the containment.
Samples of containment atmosphere were taken following the accident at Three Mile Island, but the process involved scce risk to workers and did not yield real-time information. The events clearly show a need for such information on a continuous basis following an accident.
It is essential that the operator have continuous information as to the hydrogen concentration for an indication of the need and use of reactor pressure vessei venting or containment combustible gas control systems.
1235 211
It is concluded that containment pressure, containment water level, and continuous indication of hydrogen concentration in the containment atmosphere will provide critical infonnation to the operator on containment conditions during and following an accident. These parameters should be provided in the control room of all reactor power plants.
We further note that an effort is currently underway to revise Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident."
The revision will include additional parameters that should be provided to the operator in order to assess plant conditions during the course of an accident. The list of parameters will take into account all reccmmendations, including those from the nuclear industry and the public, and will supplement those itmes reccmmended by the TMI-2 Lessons Learned Task Force.
3.
POSITICN Consistent with satisfying the requirements set forth in General Design Criterion 13 to provide the capability in the control room to ascertain containment conditions during the course of an accident, the fo! lowing requirements shall be implemented:
(1)
A continuous indication of containment pressure shall be provided in the control room. Measurement and indication capability 1235 212
. shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and minus fiv,e psig for all containments.
(2)
A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control rocm.
Measurement capability shall be provided over the range of 0 to 10% hydrogen concentration under both positive and negative ambient pressure.
(3)
A continuous indication of containment water level shall be provided in the control rocm for all plants. A narrow range instrument shall be provided for PWRs and cover the range frem the bottom to the top of the containment sump.
Also for PWRs, a wide range instrument shall be provided and cover the range from the bottom of the containment to the elevation equivalent to a 500,000 gall.on capacity. For BWRs, a wide range instrument shall be provided and cover the range from the bottcm to 5 feet above the normal water level of the suppression pool.
The containment pressure, hydrogen concentration and wide rance containment water level measurements shall meet the design and cualification provisions of Reculatory Guide 1.97, including qualification, redundancy, and testibility.
The nar ow range containment water level measurement instrumentation shall s1235 213
. be qualified to meet the requirements of Regulatory Guide 1.89 and shall be capable of being periodically tested.
l 1235 214 k
ENCLOSURE 4 INSTALLATION OF REMOTELY CPERATED HIGH POINT VENTS IN THE R 1.0 Introduction 10 CFR Part 50.46 requires that af ter any calculated successful initial cperaticn of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extented period of time Additionally, required by the long-lived radioactivity remaining in the core.
Criterion 35 of 10 CFR Part 50 Appendix A requires that a system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat frcm the reactor core folicwing any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) metal-water reacticn is limited to negligible amounts.
During the TMI-2 accident, a condition of low water level in the reactor vessel and inadequate core cooling existed and was not rectified for a long period of t,ime.,The resultant high core temperatures produced a metal-water reaction The with the subsequent production of significant amounts of hydrogen.
collection of noncondensable gases impaired natural circulation cooling Additionally, the collection of noncendensable gases limited capability.
reactor coolant pump operational capability because of coolant voids in the Even when reactor coolant pump operation was system occupied by the gases.
possible, the installed plant venting system was capable of removing the non-condensable gases only through an extremely slow process.
The purpose of tais reccmmendaticn is to provide reactor ccolant system and reactor vessel head high point vents remotely operated from the control rocm for the purocse of removing noncondensable gases collected in the system in order to allcw satisfactory long-term core cooling.
1235 215
-eu.-m
2 2.0 Discussion The collection of noncondensable gases in the reactor coolant system at TMI-2 significantly degraded natural circulation cooling capability. There is indication that these gases were predominantly hydrogen and collected at high points in the pressurizer, in the reactor vessel dome, and in the reactor coolant system piping. For other accident sequences, in adcition to hydrogen generated by metal water reaction, other noncondensible gases could be of concern. For example, nitrogen is available from PWR accumulators, and helium or other fill gases and fission gases are available from ruptured fuel elements.
Venting of the reactor coolant system was accomplished at TMI-2 through the vent located at the top of the pressurizer, and to s:me degree through the makeup tank. Neither of t;'ese paths provided expeditious venting capability unless the reactor coolant pumps were operational. Reactor coolant pump operation permitted the degassification of reactor coolant through the pressurizer spray in the steam space. As noncondensable gases were collected in the steam space of the pressurizer, they were vented through the vent located at the top of the pressuri:er. The reactor coolant pumps provided forced circulation and aided in the dispersion of the noncondensable gasr.s throughout the reactor coolant such that the flow through the makeup tank provided another vent Jath. Reactor coolant pump operation was not poss'ble for a significant period of time, however, due to voids in the reactor coolant system. These voids were probably the result of noncondensable gases as well as steam voids. Even when the reactar coolant pumps were operational, this rather slow method of venting prevented a more orderly plant cooldown.
Sir.ce continua reactor c:clant ;umo cperation cannc: ::e ass.mec durin; transients or accidents, tha capability for natural circulation cocling must 1235 216
3 in PWRs must be maintained. The addition of remotely operated nigh point reactor coolant system and reactor vessel head vents is, therefore, required so that the accumulation of ncn-condensable gases does not impair natural circulation capability.
It is recognized that SWRs provide venting capability through the use of the Automatic Depressuri:ation System (A05).
The requirements below are applicable for BWRs as well as PWRs in order to demonstrate the adequacy of any currently installed venting capability.
3.0 Position Each applicant and licensee shall install reactor coolant system and reactor vessel head high point vqnts remotely operated from the control roca. Since these vents form a part of the reacter coolant pressure boundary, the design of the vents shall conform to the requirements of Appendix A to 10 CFR Part 50 General Design Criteria.
In particular, these vents shall be safety grade, and shall satisfy the single failure critericn and the requirements of,IEEE-279 in order to ensure a low probability of inadvertent actuati n.
Eash applican't and licensee shall provide the following informatica con 'rning the design and operation of these high point vents:
1.
A description of the construction, locaticn, size, and power supply for the vents along with results of analyses of loss-of-ccolant accidents initiated by a break in the vent pipe. The results of the analyses should be demonstrated to be acceptacle in accordance with the acceptance criteria of 10 CFR 50.45.
2.
Analyses demonstrating that the direct venting of noncondensable gases with perhaps high hydrogen concentrations does not result in violation of ccebustible gas concentration limits in containment as described in 10 CFR Part 50.44,. Regulatory Guice 1.7 (Rev. 1), and Stancard Review Plan Secticn 6.2.5.
lj235 217 F
4 3.
Procedural guidelines for the operators' use of the vents. The infomation available to the operator for initiating or teminating vent usage shall be discussed.
e 3
1235 218
EtiCLOSURE 3
flUREG-0578 ERRATA 1.
Section 2.1.5.a, pace A-16, fifth line from bottcm of_g_ ace:
a Change to read, "..
25,000 SCFM (Standard Cubic Feet per Minute)..."
Reason: Editorial change.
2.
Section 2.1.5.b, oage A-2.0. first line at top of oage:
Change to read, "However, as an interim measure cending the comore-hensive lonaer term review which must be done in this regard, it is prudent to require inerting..."
Reason: Clarify intent.
3.
Table A-1, page A-25, column entitled "SWRS":
3 Delete "Shoreham(OL)"
Reason:
Plant has recembiners.
4.
Secticn 2.1.6.b, oage A-25:
Ghange title to read, " Design Review of Plant Shielding and Environ-mental Qualification of Ecuicment for Scaces/ Systems Which May Be Used 3 Post Accident Operations."
Reason: To more clearly reflect that degradation of safety equipment by radiaticn during post-accident operation is also a principal concern addressed in this section.
5.
Section 2.1.6.b, oage A-22, fourth line from bottem of oage:
Following " Regulatory Guides 1.3 and 1.4" add "(i.e., the equivalent of 50% of the core radiciodine and 100% of the core noble gas inventory are ccntained in the primary coo?'
J,..."
Reason: Clarify intent.
1235 219
. 6.
Section 2.1.8.b, 3 age A-39, paragraph 1.b:
Change to read, "f;oble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition (ALARA) concentrations to a maximum of 106 Ci/cc (Xe-133). Multiple monitors are considered to be necessary to cover the ranges of interest. The range capacity of individual monitors should overlap by a factor of ten."
Reason: To better reflect the intent of the Task Force and practical considerations regarding current state-of-the-art for low concentration effluent monitoring.
7.
Section 2.1.8.c. Dage A-41, " Position" caragraoh at bottom of oage:
Change to read, "Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel m1y be cresent during an accident."
S.
Section 2.2.l'.b, page A-49, subparagraoh 3 under DISCUSSION:
Delete the word "and" between " identified" (in the first line of the sentence) and " personnel" (in the second line of the sentence).
Reason: Typographical error.
9.
Section 2.2.2.b, cace A-58, second caragraoh of position statement:
Change to read, " Records that pertain to the as-built conditions and layout of structures, systems and ccmponents shall be stored and filed at the site and accessible to the technical support center under emergency conditions. Examples of such records include system descrip-tions, general arrangement drawings, piping and instrument diagrams, piping system iscmetrics, electrical schematics, wire and cable lists, 1235 220
, and single line electrical diagrams.
It is not the intent that all records cescribed in ANSI ?!45.2.9-1974 be stored and filed at the site and accessible to the technical support center under emers2ncy conditions; however, as stated in that standard, storage systems shall provide for accurate retr.ieval of all pertinent information without undue delay."
- 10. Table B-1. cage 3-2, footnote (b):
Change "... after July 1, 1982" to "... after July 1, 1981."
Reason: Typographical error,
- 11. Table B-1, oage B-4, item 2.1.8.b:
Change abbreviated title frca "High Range Effluent Monitor" to "High Range Radiation Monitors."
Reason: Editorial correction to make title consistent with that used in referenced discussion section.
- 12. Table 3-1, eage B-5, item relating to Section 2.2.1.b:
Change' abbreviated title frcm " Shift ~ Safety Engineer" to " Shift Technical Advisor."
Reason: Editorial correction to make title consistent with that used in referenced discussion section.
- 13. Table 3-1, footnote a, en cages B-2. 3-3, B-4, and 3-5:
Add the words, ", whichever is later." after "or prior to CL."
Reason: Clarify intent.
1235 221
EftCLOSURE 6 P:PLEMEtiTATICil CF REQUIREMEtiTS FOR OPERATIflG PLAT 1TS AND PLA1TS IN OL REVIEW Position Sect.
Abbreviated Position Implementat]on No.
Title Description Category 2.1.1 Emergency' Power Supply Complete implementa-A Requirement tion.
2.1.2 Relief and Safety Valve Submit program descrip-A Testing tion and schedule.
b Complete test program.
3y July 1981 2.1.3.a Direct Indication of Complete implementation.
A Valve Position 2.1.3.b Instrumentation for Develop procedures and Inadequate Core Cooling
& scribe existing inst.
A flew level instrument design submitted.
A Subcooling meter installed.
A New level instrument installed.
B 2.1.4,
Diverse Containment Complete implementation.
A Isolation DedicakedH C Description and imple-A Fenetration$ ontrol 2.1.5.a mentation schedule.
Complete installation.
B Category A:
Implementation cceplete by January 1, 1980, or piror to OL, a
whichever is later Category 3:
Implementation complete by January 1, 1981 bRelief and safety valve festing shall be satisfactorily ccepleted for all plants prior to receiving an cperating license after July 1, 1981.
1235 222
2 2P'.EMENTATIONTABLE_(Continued)
Position S e ct..
ecoreviatea Positicn Implementatjen No.
Title Descriotion Catecory 2.1.5.c Recombiners Review procedures and A
bases for recombiner use.
2.1.6.a Systems Integrity for 1 mediate leak A
High Radioactivity reduction program.
Preventive maintenance A
program.
2.1.5.b Plant %ielding Review Complete the design A
revi ew.
Implement plant modifications.
6 Category A:
Implementation cceplete by January 1,1980, or prior to OL, a
whichever is later.
Category 3:
Implementation complete by January 1, 1981 1235 223
3 IMPLEMENTATION TABLE _(Continued)
Position Sect.
Abb'eviateo rosition Implementatjon No.
T..le Descrioticn Category 2.1.7.a Auto initiaticn of Complete imolementaticn A
Auxiliary Feed of centrol grade.
Complete imolementation 8
of safety grade 2.1.7.b Auxiliary Feed Flow Complete imolementati,n A
Indication 2.1.8.a Post Accident Sampling Design review complete.
A Preparation of A
revised procedures.
Implement plant modifications.
B Description of proposed modification.
A 2.1.8.b High' Range Radiation Installation complete.
B Monitors.
2 1 8.c Improved Icdine Complete implementation A
Instrumentation i
2.1.9 Transient & Accident Complete analyses, Analysis procedures and training Containc.ent Pressure Installation complete 3
Monitor ContainmentWaterIevel Installation cceplete B
Monitor Containment Hydrogen Installaticn cceplete B
Monitor RCS Venting Design submitted A
Installation ccmplete B
' Category A:
lmplementation ccmolete by January 1, 1980, or prior to CL, v.nicne.er :s latar.
Cate;cr;. S:
Implementa:icn ccmplete by January 1, 1931.
- Analyses, procedural changes, and operating training shall be provided by all cperating plant licensees and applicants for cperating licenses r'ollowing tne attacned senecule.
T235 224
4 IMPLEMENTATION TABLE _(Continued)
Position Sect.
Aboreviacec Position Implementatjen No.
Title Descriotion Category 2.2.1.a Shift Supervisor Complete implementation.
A Responsibilities 2.2.1.b Shift Technical Advisor Shift technical advisor A
on duty.
Complete training.
B 2.2.1.c Shift Turnover Complete implementation.
A Procedures 2.2.2.a Control Room Access Ccmplete implementation A
Control 2.2.2.b Onsite Technical Establish center.
A Support Center 2.2.2.c Onsite Operational Ccmplete implementation A
Support Center "Categcry A:
Implementation complete by January 1, 1980, or prior to OL, whichever is later.
Category 3:, Implementation cceplete by January 1, 1981.
1235 225
5 ANALYSIS AND TRAINING SCHEDULE Task Descriotion Completion Date 1.
Small Break LOCA analysis and preparation of emergency procedure guidelines July-September 1979*
2.
Implementation of small break LOCA emergency procedures and retraining of operators December 31, 1979 3.
Analysis of inadegaate core cooling and preparation of emergency procedure guidelines October 1979 4.
Implementation of emergency procedures and retraining related to inadequate core cooling January 1980 5.
Analysis of accidents and transients and preparation of emergency procedure guidelines Early 1980 6.
Implementation of emergency procedures 3 months after and retraining related to accidents guidelines established and trairisients 7.
Analysis cf LOFT small break tests Pretest (Mid-September 1979)
" Range covers completion dates for the four NSSS vencors 1235 226