ML19260B141
| ML19260B141 | |
| Person / Time | |
|---|---|
| Issue date: | 08/07/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1646, NUDOCS 7912070296 | |
| Download: ML19260B141 (81) | |
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MEETING DATE: 6/13/79 ISSUE DATE:
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/O / f. l':t MINUTES OF THE ACRS SUBCOMMITTEE '4EETING ON REACTOR OPERATIONS WASHINGTON, DC
. lune 13',1979 On June 13, 1979, the ACRS Reactor Operations Subcomittee met in Washington, DC, to discuss the Millstt.ne 2 power level increase and the NRC Safety Research Pro-gran on Operational Safety. Notice of this meeting appeared in the Federal Register on May 29,1979.
Attachment A is a copy of the meeting agenda. The attendance list is in Attachment B.
Attachment C is the tentative schedule for the meeting. Selected handouts received at the meeting are in Attachments D through G of these minutes. A complete set of the handouts is attached to the office copy of these minutes.
The Subcomittee had received no written comments or requests for time to make statements from members of the public.
MILLST0flE 2 POWER LEVEL INCREASE INTRODUCTORY STATEMErlT Mr. H. Etherington, Chairman of the Reactor Operations Subcommittee, called the meeting to order. Another member present was Mr. W. Mathis. He stated the purpose of the meeting and said that the morning session will be concerned with the Millstone 2 power level increase.
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Reactor Operations 6/13/79 Millstone 2 was originally reviewed by the ACRS for operation up to 2570 MWt a*d the present request is iar operation up to 2700 MWt. A review of this procesal is referred ta,the Reactor Operation Subcommittee instead of the Millstone Subcommittee in accordance with the Conmittee procedure expressed in a letter from Mr. Fraley to Mr. Gossick dated May 12, 1978. This letter was read for the records. An important paragraph from the letter follows:
"It is our understanding the proposals to extend operating power levels beyond that originally established as the design power level will normally involve the formal ACRS review and report."
The review, Mr. Etherington continued, would be held, by the Millstone 2 Subcom-mittee.
This Subcommittee will allot appropriate ti.ne to the review of a simple case of pcwer increase to the original review level. But, Mr. Etnerington questions whether this case falls in that category or whether this Subcommittee is the proper one to review this application.
Mr. Etherington stated that a review of the Staff. valuations suggest that although Northeast Utilities designed the plant for an ultimate operation at 2700 MWt, there was only a limited analysis for that power level. Moreover, the present analysis justifying operations for 2710 MWt is largely based on CE topical reports and codes that did not exist when the FSAR was resiewed.
Therefore, Mr. Etherington asked the Staff to address the following questions:
1.
What is the Staff basis for considering Millstone 2 stretch power that has been reviewed by ACRS?
2.
Could the power increase have been supported on the basis of the original SAR methodology?
3.
Does the new methodology lead to a higher power level than the original basis?
4.
Which CE plants have been reviewed at the POL stage by the Staff and ACRS on the basis of the new methodology?
5.
What are the pertinent topical report numbers, and supply any infonnation on ACRS subcommittee or Committee reviews of topical reports?
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Reactor Operations 6/13/79 Mr. Conner, NRC Staff, addressed the Subcommittee ccacerns.
W;th regards to question 1, he stated that the FSAR is based on 2700 MWt and all but a few of the accidents were analyzed at 2700 MWt, however, the Staff SER approved 2560 MWt because that was the power level the licensee app'.ied for. The Environmental Statement was completed for 2700 MWt.
Mr. Conner stated that the MAR is exactly the same as Calvert Cli#fs, which has already gone to stretr',0wer.
Millstone was reviewed after Calver Cliffs 1, and before Calvert Cliffs 2.
Mr. Etherington then stated that if, if fact, Calvert Cli#fs was raised to the stretch power on the same basis as being requested here, I think we could easily get the full Committee to approve the stretch power increase.
Mr. Etherington read from the Calvert Cliffs letter the following:
"Each reactor is designed to produce 2440 MWt with a stretch power to 2700 Mut."
In other words, Mr. Etherington stated that, at the time the Committee reviewed this report on March 13, 1969, the plant had, in fact, been designed for the higher power and had been analyzed for the higher power, and Calvert Cliffs is not a precedent.
The Chairman suggested that we complete the Millstone review as scheduled and let the full Committee decide on an acceptable basis for the review of stretch power plants.
Concerning question 2, Mr. Conner stated that Millstone 2 could operate at 2700 MWt on the basis of the old methodology reported in the FSAR. However, there would not be adequate margins to trip points. The new methodologies pro-vide better margins, more stable operation, and fewer spurious trip. This con-cern was experienced on Calvert Cliffs I when they received a stretch power license and found that they did not have adequate margins and that they have to adopt calculational techniques developed since the stretch power was approved This to provide operating flexibility so that they could run at 2700 MWt.
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, 6/13/79 Reactor Operations change was used for Cycle 2 reload, wnere tney used TORC /CE-1 methodology to obtain new trip thermal and pressure margins.
Mr. Hart of Northeast Jt'ilities Service Company (NUSCo) presented a trief description of the plant and site. He also presented the licensing and operating history of Millstone 2.
Following are highlights of his presentation:
Acolication et construction pennit, February 1969 Issuarte of construction permit, December 1974 Limited work authorization, November 1969 FSAP submitted, August 1972 Operating license issued, August 2,1975 Initial core assembly started the day after the issuance of the operating license. Completed August 10, 1975.
Initial criticality, October 1975 Commerical operation, December 26, 1975 Diesel generator replaced, December 20, 1975
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Main condenser retubed, May 1977 First refueling completed, April 1977 During the outage extensive steam generator plugging was performed.
Overall capacity for Cycle 1 was 66.6 percent Cycle 2 initial criticality, April 21, 1978 -- Capacity for Cycle 2 was 94.4 percent.
Current fuel cycle reached criticality on May 18, 1979, and reached 2560 MWt on May 31,1979 Mr. E. C. Farrell of NJSCo. next discussed the problem of the two power operated relief valves which are the cause of the current plant outagc. In the Millstone 2.
configuration there are two isolation valves which isolate the power ope atna relief valves. After startup, a small seal ring leakage from one of the isolation valves for the' power operated relief valves exceeded the Tech. Spec. limit of 1 gpm 1$11 129
, 6/13/79 Reactor Operations unidentified leakage in the RCS. The plant was shut down and the seal replaced.
This has been a recurring problem with this valve.
Mr. Hart then continued with the presentation on the overview of the power increase. The followinq are highlights of his presentation:
The NSSS had an excess capacity in order to assure the 2560 MWt power level.
From experience CE gained from Palisades and Maine Yankee, it was determined that a stretch capability of 2700 MWt was technically feasible.
On this basis, NUSCo. and Becthel designed all the equipment for 2700 MWt.
Therefore, the plant was designed, built, and the'FSAR was written to accommodate operation at 2700 MWt. The plant wa's conservatively licensed at 2560 MWt.
On December 15, 1978, NUSCo. filed an application for 2700 MWt. This additional increase saves approximately 400,000 barrels of oil per year.
The value of this additional power is estimated to be worth about $4 million in the first year assuming 65 percent capacity.
Most of the Chapter 14 accidents needed to be reanalyzed.
Review confirmed that no new hardware changes were required.
Based on the reanalysis, the following items were required:
1.
Credit was taken for the charging pumps to mitigate a small break LOCA.
- 2. 'In order to increase the margin for the four pump loss of flow trip an electronic reactor coolant pump speed sensing system was installed.
Mr. Hart stated that a shield was installed around the reactor flange during 1979 refueling outage in order to attenuate the effects of neutron streaming from the annulus area between the reactor vessel and the primary shield. Containment entry will now be less restrictive. This change is not stretch-power related. Another non stretch-power related change is the increase in the cold leg temperature from 542 F to 549 F.
This has resulted in an increased turbine inlet pressure which has had a significant increase in turbine generator output due to cycle efficiency increase. A small amount of electrical '
power was also recovered due to improved moisture separator reheater efficiency.
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- 6/13/79 Reactor Operations Mr. Harris of NUSCo. stated, in answer to a question, that the increase in heat transfer area comes from the insertion of fuel rod assemblies that do not contain fixed, burnable shims, but by rods containing baron, carbide, and aluminum oxide.
Mr. Harris next discussed power increase methodology changes. He stated that with the exception of the item discussed previously on small break LOCA, Millstone 2 could obtain the 2700 MWt rating without use of any new methodology. Thus, as part of the safety analysit portion of the effort of stretch power, NUSCo.
used new methodology and availed themselves the opQortunity to make changes in certain inputs, and to take credit for selective systems. The following summarizes some of these key changes:
1.
Used TORC /CE-1 multi-channel themal hydraulic code.
2.
Used CEFLASH-4AS to analyze small break LOCAs.
3.
Used statistical combination of uncertainties in inputs used in setpoints and the detemination of themal margin limits. This results in a 5 percent credit compared to previous methods. For Millstone 2 a partial credit of 3 percent was conservatively taken.
Mr. Harris then discussed changes in key input assumptions and assumed systems taking credit for the safety analysis supporting 2700 MWt operation.
Increase the hot scram times from 2.75 to 3.1 seconds.
Changing inlet temperature from 542 to 549 F.
Assumed uncertainties in measured power distributions. Fg and Fg used 6 and 7 percent, respectively, as compared to 7 and 8 percent used for Cycle 2.
Reactor coolant pump speed sensing system tas been changed which results in lower uncertainties and faster response time compared to the steam generator delta p system for a low flow trip. Credit was taken for this system on tr. four pump loss of flow event, which is the DNBR limiting transient for Millstone 2.
Mr. Gurney of NUSCo. next discussed the result of the transients analysis performed in npport of stretch power licens,ing. The major analysis change was in the use of iORC/CE-1 methodology. In addition, two other changes were accounted for in the analysis.
1.
Use of reduced flow guide tube design increased the safety analysis scram time from 2.75 to 3.1 seconds.
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, 6/13/79 Reactor Operations 2.
Use of the reactor coolant pump speed sensing system which changed the four pump loss of flow setpoint from 89 percent flow to 9.5 percent.
Mr. Gurney discussed the transients reanalyzed and presented the results.
Mr. Hart discussed the ACRS generic list in Report No. 7 dated March 21, 1979. Their reviewers have determined that there were four items which are sensitive to power level increase, albeit small due to a small 5 percent increase. The affected items are as follows:
1.
Effective operation of containment space in a LOCA.
Behavior of reactor fuel under abnormal conditio'ns.
2.
3.
Steam generator tube leakage.
4 Vessel support structures.
All the above items have been addressed by NUSCo. and found satisfactory by the Subcommittee.
Mr. Conner then started the NRC presentation. He gave a brief summary of the chronology of the licensing process for the Millstone 2 stretch power appli-cation. Mr. Conner then gave a statement of the Staff position regarding the Three Mile Island 2 accident and how it affects the power level increase. He stated that for a small power increase such as 5 percent for Millstone, the Staff would not require that the review include matters as a result of the Three Mi.le Island 2 accident except for matters raised by I&E Bulletin 06B.
Mr. Conner then presented a list of facilities with stretch power increases.
He cited as an example, Maine Yankee, which first had a licensed power level of 2440 MWT and an FSAR power level of 2550 MWt. They then came to the ACRS and after modification, they have gone now to 2630 MWt. This is above their original FSAR. In Calvert Cliffs, this was not done. The FSAR ultimate level was 2700 MWt and that is the power level they have obtained by going through the process of power increase. In Millstone 2, 2560 MWt and 2700 MWt was the FSAR power level and this is what is now pending.
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Reactor Operations 6/13/79 In reply to Mr. Etherington's question, Mr. R. Mills of CE stated that he will provide a listing of the accidents analyzed and the techniques used and available at that time for the full Committee.
In Millstone 2, Mr. Conner stated that the core thermal output was analyzed at 2560 MWt, and the site parameters at 2700 MWt. All major equipment including the ECCS, the containment, and certain postulated accidents, were designed at 'the 2700 MWt level. These will be identified for the full Commi ttee. The original SER was performed at 2560 MWt, realizing that many of the input parameters were for 2700 MWt, but the FSAR says it's for 2560 MWt.
However, the FES is for 2700 MWt. He added that in the Calvert Cliffs case the power levels used in the SER and FES were reversed, i.e., the Staff evalua-tion in the SER was for 2700 MWt and the FES was for 2560 MWt.
The Staff has reviewed the Applicant's proposal for a power level increase and found that conclusions reached in the FES and subsequent EIS issued since that time remain valid.
Mr. Etherington during caucus stated that he has no problem for an increase in power level on a technical basis, but questions the methods used in the licensing process.
The Subcommittee recomended that the full Comittee review this application and a tentative agenda for the full Committee meeting was discussed with the Staff and Applicant.
THREE MILE ISLAND 2 IMPLICATION TOPICS Mr. Etherington continued the meeting in executive session and considered the list the Mr. R. Fraley made concerning topics the Committee had expressed interest in and assigned them tentatively to various existing and new Sub-comittees. Mr. Etherington stated that the document has no standing; it has
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Reactor Operations 6/13/79 not been reviewed or adopted, but there were only two topics of interest, Maintenance QA and Security, and Alignment of Systems for Maintenance Testing.
He felt that discussing these items now ma i prevent another meeting.
After a brief discussion, the Subcommittee tabled the discussion after deciding that they would discuss it further at some opportune time and to wait until Fraley's document was acted upon and approved by the Committee.
NRC SAFETY RESEARCH PROGRAM Mr. Etherington quoted from a document by the House' Committee on Interior and Insular Affairs. It states in essence that in order to increase the use of the ACRS research report it should do the following:
1.
Prepare a clear statement of priorities.
2.
Prepare its report with a schedule that assists the Commission in preparing the FY 81 budget request.
3.
Include discussion in which the Commission's reactor safety research projects are expected to affect the Commission's Reactor Regulations.
In light of this direction, the ACRS plans to write a letter to the Commission describing the extent to which the proposed FY 81 budget is responsive to recommendations by the ACRS in its 1978 report to Congress, and possibly commenting on other items in the proposed budget.
Mr. Etherington stated that the section relevent to the annual report to Congress is Section 15, Operational Safety. The current Safety Research Program referenced in our 1978 report were as follows:
Evaluation of qualification testing Fire Protection Noise Diagnostics Man-Machine Interfaces Mr. Etherington then read the recommendations from the 1978 report on Section 15.
Dr. G. Bennett, NRC Staff, discussed the operational safety research program.
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. 6/13/79 Reactor Operations Fire Protection:
The fire protection program was. designed to provide an improved and independent base for the evaluation of fire protection testing methodology and also in the preparation of regulatgry guides and standards.
Four organizations are invohed. They are: Sandia Laboratories, Underwriters Laboratories, Applied Physics Laboratory, and IEEE. Dr. Bennett then discussed the program each organization performed. Attachment D provides details of the program.
qualification Testino Evaluation Procram:
The objectives of this program is to provide an independent assessment of LOCA and other testing methodologies. Sandia is the principal contractor for this work. Attachment E provides details of this program.
Mr. V. Benaroya, NRC Staff, next presented the technical assistance program on fire protection. During 1978, $180,000 were spent to support the fire pro-tection reviews of 37 applicants. The TAP was with Brookhaven National Labs.,
which in turn has subcontracted to Babcock Associates in Chicago. For FY 79, a comoetitive contract was let to Babcock Associates for $180,000. For FY 80 another $115,000 ill be needed. These contracts are for extension of the branch manpower.
Mr. Satterfield, NRC Staff, also has a similar program in the equipment quali-fication area. He is supporting a two man-year level of effort at the Oak Ridge National Lab. in assisting with the equipment qualification reviews.
Funding effort is about $125,000 for this year.
Mr. R. Feit, NRC Staff, briefly described his visit to Europe and Japan concerning fire protection and qualification testing evaluation. He stated that they generally follow the US lead although they are unhappy,witr the procedures and approaches of some of our testing programs. They are now leaning in the direction of guaranteeing that the equipment will work so they can market it.
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. 6/13/79 Reactor Operations I n the area of fire protection. the French, it was stated, have an elaborate program. They are spending about $50,000 a year on basic combustion chemistry.
One avenue of research,i's the use of proper ventilation to controi fire. They have looked at halon as a possibility and have rejected it even though other European countries are sold on it. The French have rejected halon because of its toxicity. Mr. Benaroya stated that the 1211 halon are more toxic than our 1301 halon. He also stated that the 1211 cost less than the 1301 but the total cost is negligible for the whole plant.
Dr. Bennett continued his presentation in the opera:Ional safety program.
Noise Diaanostics:
This program is set up in cooperation by NRR and Oak Ridge to provide experi-mental and anal tical studies of abnormal reactor behavior using noise diag-
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nostics and also to develop criteria for loose parts monitoring system assess-ments. In addition to this program, Oak Ridge is providing technical assistance to NRR in this area. NRR has asked Oak Ridge to participate in an ASME Committee on piping vibration monitoring standards. Further details of this program are found in Attachment F.
Recently, the Staff requested that the noise diagnostic system at Oak Ridge be transferred to TMI-1. They want to have the system operational through the next cycle.
The sy!, tem was also used during the height of the TMI-2 crisis. The system on TMI-2 indicated that no loose parts could be detected, no boiling could be detected at the core exist, and that the core barrel configuration appeared to be normal. The system was also used to verify natural circulation. The temperature noise system indicated that no bulk core boiling could be detected.
All the signals used as input for the system were obtained from the Metropolitan.
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Reactor Operations 6/13/79 Some foreign work in noise diagnostics is being performed in Sweden, Japan, and Gemany. Oak Ridge is monitoring this program.
Human Factors:
The objectives of this program are to assess the role of human errors in reactor safety, and study ways in which human errors can be reduced. The principal research organizations are Oak Ridge, Idaho, and Sandia. Sandia is involved as a consultant capacity. The principal activity underway at Oak Ridge involves the safety-related operator action study. User of this program is the Office of Standard Development.
The,, purpose of this pro-gram is to establish a data base to assess proposed criteria, specifically ANSI Standard N-660 for determining whether or not safety-related operator actions have to be automated. More detail of the human factors programs is presented in Attachment G.
The folicwing programs for 1981 were discussed:
Fire protection to concentrate on fire suppression.
Qualification testing _ post mortem work on TMI-2.
Human factors -to focus on simulator studies and control room design.
Noise diagnostics-to continue work on baseline plant signatures.
Safety and relief valve testing - this program is still in the dis-cussion stage with NRR.
New prog' rams - proposing evaluation of computerize control systems.
Also looking into alternate control systems such as fluidics, fiberoptics, etc.
Inspection and Enforcement - provide support in the operational safety area to develop methods for improving assessment of licensee performance.
Reactor Operational Safety - studies to review all operational aspects of nuclear power plant operation and to develop a more systematic research program on operational safety topics, rather than the almost fire drill aoproach now being used.
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. 6/13/79 Reactor Operations A supplemental budget of $500,000 on relief valve testing was discussed.
After the presentation by Dr. Bennett, a brief caucus was held. Mr. Etherington inquired if the recommindations made in the last ACRS report to Congress were responded to satisfactorily in the research program being performed and planned.
It was the consensus of the Subcommittee that the recommendations were responded to satisfactorily.
Mr. Etherington thanked Dr. Bennett for the fine presentation and adjourned the meeting at 4:40 p.m.
NOTE: For additional details, a complete transcript of the meeting is available in the 11RC Public Document Room,1717 H Street, NW, Wasaington, DC 20555, or from Ace-Federal Reporters, Inc.,
444 North Capitol Street, NW, Washington, DC.
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1 Federal Register / Vol. 44. No.104 / Tuesday. May 29. 1979 / Notices 30*/85 Director for Nuclear Reactor Regulation consultants to verify the piping stress The agenda for subject ineeting shall concluded that the public health and reanalysis.
be as follows:
safety required that the affected In addition to revi' dig the licensees
- Wednesday./une 13. Jers facilities be placed in a cold shutdown corrective actions the NRC is reviewing N "' ""# M ' ##"'l"'i'" # @ ##"' "-
condition pending further order of the any generic implications at other Commission. Orders to this effect were facilities.The NRC's Office of Inspection ne Subcommittee may meet in Execunve issued to the licenseca of the above and Enforcement !ssued Information Session, with any ofits consultants who may Notice (IN) No. 79 06, which described be present. to explore and exciange their reactors.
mjte w
p onsr ne Orducs provide that within 20 the event, on March 23,1979, to aH days each licenses must respond witk holders of reactor operating licenses and is formulate a -ort and recommendations respect to:
construction permits. On April 14.1979, to tne ruu Committee.
the NRC's Office ofInspection end At the conclusion of the Executive Sassion.
(1) why the licensee should not Enforcement issued Bulletin No. 79 07 to the Subcommittee will hear present tions by reanalyze the facility piping systems for applicable licensees which identified and hold discussions with representatives of seismic loads on the piping system and actions to be taken. This includes the NRC Staff. the Northeast Nuclear Energy any other affected safety systems using identification of the methods of analyses any NhEC d th co ts; an appropriate piping analysis co nputer used, how they were verified. safety may then caucus to determine whether the code which does not combine loads systems affected. and a plan of action t matters identified in the imtial session have algebraically, assure plant safety. As of May 9,1979.
been adequately covered and whether the (2) why the licensee should not make the NRC has received responses to project is ready for review by the full any modifications to the facility piping Bulletin No. 79-07 from alllicensees of Committee.
systems indicated by the reanalysis, and operating reactors except for nree Mile In addition. it may be necessary for (3) why facility operation should not Island Units 1 and 2 which are the Subcommittee to hold one or more continue to be suspended until shutdown. ne NRC staff is reviewing completion of the reanalysis and any these responses on a high prianty basis.
closed sessions for the purpose of required modifications.
Additional actions will be taken as exploring matters involymg propnetary Information. I have determined. in All of the pants are now in a cold appropriate.
accordance with Sub.W.10(d) of shutdown condition. (Surry Unit 2 was Weehmston. D.C. thia 14th day Pub. L 92-483, that, shoula such already in an extended outage for steam [ ate sessions be required. it is necessary to generator replacement.)
g Cause or Causes-The uncertainty m close these sessions to protect Samuel). chi!k.
proprietary information (5 U.S.C.
the calculated piping stresses and S*CT#'Y W' C#8"#8'0"-
552b(c)(4)).
support loadings in safety-related piping systems at the five plants is attnbutable PD= * " N 5*-* " =1 Further information regarding topics s u a coca mus * '
to be discussed, whether the meeting to the incorrect application of the algebraic summation technique in the has been cancelled or rescheduled. the Chairman's ruling on requests for the SHOCK 2 subroutine of the P1PESTRESS A
Re opportunity to present oral statements computer code. proprietary to Stone and Saf ards, S@ committee on and the time allotted therefor can be Webster.
Operating Reactors; Meeting obtained by a prepaid telephone call to Actions Taken to Prevent Recurrence The ACRS Subcommittee on the Designated Federst Employee for Licensee /Arbitect Engineer" Operating Reactors will held a meeting this meeting. Mr. Richard K. Major.
Identification of all safety related on June 13.1979, in Room 1048.1717 H (telephone 202/e34-1414) between 8:15 systems that have been analyzed with a Street. NW., Washington. DC 20555 to a.m. and 5 00 p.m EDT.
piping computer code involvmg a review a request for a power level Background infor. nation concerning program deficiency is underway.
increase at the Millstone Nuclear Power items to be considered at this meeting Computer inputs are being checked to Station. Unit 2. and to discuss the 1979 can be found in documents on file and assure that all reanalyzed pipmg will Review and Evaluation of the NRC available for public inspection at the reflect the as-built condition at each Safety Research Program on NRC Puvlic Document Room.1717 H plant. Piping analyses are being rerun Operational Safety. Notice of this Street. NW Washington. DC 20555 and, and piping and supports exceeding meeting was published in the Federal regarding the Millstone Station at the allowable stresses will be identified.
Register on May 24.1979.
Waterford Public Library. Route 158.
Modifications Mll be made as In accordance with the procedures Rope Ferry Road. Waterford CT 06285 necessary, outlined in the Federal Registar on Dated. May 22,1979' MIC-ne NRC ordered each of the October 4.1978 (43 FR 45928). oral or N "'Y *'
I utilities of the five identified nuclear written statements may be presented by Advisory Coaumtree Management oficer.
power plants to shutdown their plants members of the public, recordings will P Da **5*' N * * " =1 within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The utilities are to be permitted only during those portions s e a coca ne m remain shutdown pending further order of the meeting when a transcript is being of the Commission. De NRC is in kept. and questions may be asked only contact with the licensees and the by members of the Subcommittee,its Advisory Committee on Reactor architect engineer on, actions being consultants, and Staff. Persons desiring Safeguards, Subcommittee on taken. Piping stress computer codes to to make oral statements should notify 8
be used for reanalysis of the piping wdl the Designated Federal Employee as far d
e be tested with NRC established m advance as practicable so that benchmark problems. Also, an appropriate arrangements can be made An addition, noted below, has been independent audit of selected piping ta allow the necessary time during the made to the agenda of the June 2.1979 nms will be conducted by NRC meeting for such statements.
meeting of the ACRS Subcommittee on Amcevr A
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ATTENDEES LIST ACRS SUBCO:iMITTEE !!EETING ON REACTOR OPERATIONS MILLSTONE 2 POWER LEVEL INCREASE WASHINGTON, DC JUNE 13,1979 ACRS NRC STAFF H. Etherington, Chairman M. Conner W. Mathis J. Shedlosky, Region I E. Igne, Designated Federal Dnployee P. Kapo S. Weiss G. Bennett NORTHEAST UTILITIES V. Benaroya M. Stolzenberg R. Hart G. Rhee E. Foster H. Scott E. Farrell R. Satterfield R. Kacich R. Smith J. Kelley R. Feit R. Rodgers P. Gurney COMBUSTION ENGINEERING A. Roby R. Harris R. Mills J. Bahr EXXON NUCLEAR C0.
F. Carpentino OHtHA PUBLIC POWER DIST.
S. Jensen G. Owsley K. Morris J. Gasper FLORIDA POWER & LIGHT CO.
T. Grozan PUBLIC A. Rilef J. Mulligan L. Klein M. Bloom D. Hoffman G. Womack K. Ota R. Mitchell M. Meltzer ATTAc.mtm,b 1511 140
TENTATIVE SCHEDULE ACRS SUBCOMMITTEE MFETING ON REACTOR OPERATIONS PCMER LEVEL INCREASE WASHINGTON, DC JUNE 13, 1979 MILLSTONE 2 - POWER LEVEL INCREASE APPROXIMATE TIME I.
EXECUTIVE SESSION (OPEN) 8:30 am
- 12. LICENSEE PRESE? CATION 8:45 am A.
PIANT AND SITE DESCRIPTION B.
LICENSING AND OPERATING HISTORY C.
CVERVIEW ON POWER INCREASE D.
CYCLE 3 CORE DESIGN E.
POWER INCREASE METHODOLOGY CHANGES F.
'IRANSIE?C/ ACCIDENT ANALYSIS G.
CiCLE 3/ POWER INCREASE MODIFICATIONS 1.
Credit taken for charging pump flow in IDCA analysis 2.
RPS trip from a lCP speed sensing signal 3.
Neutron shield H.
ACRS GENERIC LIST
- BREAK ************************
10:4' - 10:55 am III. NRC PRESENTATION 10:55 am A.
DERODUCTION B.
ACRS GL Lt.Trt.x AND LICENSING CONDITION ISSUES C.
CYCLE 3/ POWER INCREASE SAFETY EVALUATICN 1.
Staff review of IID, E.F.G.
2.
CEA guide tube integrity 1511 14,1 3.
Steam generator surveillance 4.
Piping system reanalysis ATTAC.uMENT C
REAC'IOR OPERATIONS / POWER LEVEL INCREASE JUNE 13, 1979 MILLSTONE 2 - POWER LEVEL INCREASE APPROXIMATE TIME D.
GENERIC ITEMS RELATED 70 POWER L.rVEL INCREASE E.
REGULA70Rh EXPERIENCES 1.
LERs 2.
Abnormal occurrence 3.
Radiation exposure history F.
WI-2 BULLETIN STATUS 11:45 am IV.
CAUCUS V.
MEETING WITH WE LICENSEE AND STAFF 12:00 pm VI.
AIDOURNMENT
- LUNCH -
12:00 - 1:00 pm RESEARCH REPORT TO CONGRESS, SECTION 5 I.
EXECtTTIVE SESSION 1:00 pm II.
? I-2 IMPLICATION, ACRS TOPICS TCPIC 10 - MAINTENANCE QA:
SECURING AND REALIGNMENT CF SYSTEMS FOR MAINTENANCE TESTING TCPIC 12 - PIANT MANAGEMENT STRUCTUR.e (NCfrE:
IF W E STAFF IS NOT READY, THIS MATTER WILL BE DISCUSSED IN EXECtfrIVE SESSION)
III. NRC STAFF PRESENTATION 1:45 pm A.
EOR LETTER TO CCNMISSIONERS, STAFF WILL
- DESCRIBE WE EXTENT 'IO WHICH WE FROPCSED FY 81 BUIX3ET IS RESPONSIVE TO RECCMMENCATIONS BY ACRS IN ITS 1978 REPORT 10 CCtCRESS E
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MILLES'IONE 2 TENTATIVE SCHEDULE JUNE 13, 1979 APPROXIMATE TIME RESEARCH REPORT TO CONGRESS, SECTION 5 COMMENT ON CTHER ITEMS IN 'IEE PROPOSED BUIX3ET-B.
FOR RESEARCH REPORT, STAFF WILL DISCUSS:
ITS FY 1978 BUDGET PPOPOSALS VIS-A-VIS ITS ACRS RECOMMENDATIONS
- ITS CURRENT PROGRAM PARTICIPATION IN NEW RESEARCH ITEMS RESULTING FROM THE 'D4I ACCIDENT C.
TECHNICAL ASSISTANCE PROGRAM
- FIRE PROTECTION V. Benaroya, DSS D. Notley, OSD
- QUALIFICATION TESTING R. Satterfield, DSS
- BREAK *******************
4: 30 pm - 4: 40 pm IV.
CAUCUS AND DISCUSSION WITH STAFF 4:40 pm 5:00 pm V.
ANOURNMENT 1511 143
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's RESEARCil SUPPORT BRANCil FIRE PROTECTION RESEARCll
SUMMARY
OF PRit!CIPAL ACCOMPLISilENTS
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INDUSTRY SURVEY CABLE SCREENING TESTS EVALUATION OF ADE0llACY OF R.G.1.75 CABLE TRAY SPACING FOR ELECTRICALLY INITIATED FIRES IN IEEE-383 00ALIFIED CABLE EVALUATION OF ADE0VACY OF R.G.1.75 CAllLE TRAY SPACING lil A STACKED TRAY CONFIGURATION FOR A PROPANE FUELED EXPOSilRE FIRE EXPEP.iMENTAL INVESTIGATION OF Tile ADEQUACY OF FIRE RETARDANT C0ATINGS AND FIRE SillELDS DEVELOPMENT OF PARAMETERS FOR D01:0R TRAY FIRE lli VERTICAL ORIENTATION i.b'-
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PRELIMINARY TESTS OF FIRE SilPPRESSION FACILITY OBJECTIVES VERIFY PROPER E0lllPMENT OPERATIONS VERIFYFUNCTIONABilllYOFDATAACQUISITIONSYSTEM FAMILIARIZE PERS0llflEL WITil GPERATING PROCEDURES SCllEDULE FACILITY IS COMPLETED CllECK00T TESTS COMPLETE - JllNE 29, 1979 DOCUMENTATION OF CilECK0llT TESTS - JULY 31, 1979 TEST PLAN SilBMITTAL - JllLY 1979
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'J FIRE PROJECTION RESEARCil PROGRAM REPLICM10fLIESIS NRR RESEARCil REQUEST - JAtlllARY 19, 1979 PERFORM CONFIRMATORY FIRE TESTING OF DETAILED MOCK-UPS OF SELECTED CONFIGURATIONS WITil FIRE PROTECTION MEASllRES I)ESIGNED IN ACCORDqt CE W PROTECTION GUIDELINES ANI) F00 TID ACCEPTABLE BY Tile NRR STAFF.
Pl!RPOSE TO C0tlFIRM WilETilER TilESE FIRE PROTECTION C0t!FIGURAT10NS ARE VAllD FOR OP PLANT CONFIGURATIONS 2.
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FIRE PROTECTION RESEARCil PROGRAM REPLICATION TESTS RANCl10 SECO MAKEUP PUMP ROOM CONSERVATIVE LARGE OIL llAZARD AltKANSAS Oi!E, llNIT 1 - AUXILI ARY BUILDING IIALLilAY C0liSERVATIVE TYPICAL llALLWAY BRllNSHICK lilTAKE STRUCTURE BASEMEllT INTERMEDIATE TEST OF FIRE SUPPRESSI0fl FOR SAFE SilUIDOWN CABLES BROWNS FERRY llNITS 1 All0 2 - REACTOR BUILDING ACCEPTABLE lilGil CONCENTRATION OF CABLE EXIRA ARKANSAS ONE, llNil 1 - CABLE SPREADit!G ROOM ACCEPTABLE DIRECTED WATER SPRAY
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FIRE PROTECTION RESEARCil PROGRAM REPLICATION TESTS TYPES OF TESTS OPERATION OF Tile SilPPRESS10ll SYSTEMS AS INSTALLED, WITil REALISTIC RESPONSE PARAMETERS PURPOSE:
CONFIRMATION OF ADE0llACY OF l'ERFORt1At!CE OF INSTALLED FIRE',
SUPPRESSION SYSTEM DISA13 LED FIRE SUPPRESSION SYSTEM WITli SulTABLE ACTION OF WilAT MIGill RE BE EXPECTED OF A FIRE IlRIGADE ACCEPTANCE CRITERIA CIRC'JIT CONillfulTY CRITERIA ESTABLISilED FOR TilAT TEST J10CK-UP MUST BE SATISFI FOR 110111 TEST RUNS.
CRITERIA WILL GENERALLY INVOLVE MAINTAINit!G CIRCulT INTEGRITY OF SUFFICIEllT DIVIS10tlS TO EFFECT SAFE SiluTDOWil
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FY 1980 PROGRAM ASSUMPTIONS PROGRAM:
WRSR TITLE:
FIN NO:
A1010 FIRE PROTECTION RESEARCH CONTRACTOR:
SANDIA SITE:
ALBUQUERQUE STATE:
NEW MEXICO NRC TECHNICAL MONITOR:
RONALD FEIT PRINCIPA' INVESTIGATOR:
LEO KLAMERUS OBJECTIVES:
TO PROVIDE INDEPENDENT DATA NEEDED IN SUPPORT OF NRC FIRE PROTECTION REQUIREMENTS OR TO MODIFY THEM AS APPROPRIATE CONCERNING:
(1 ) CA5tE AND CABLE TRAY CO}iFIGURATION DESIGNS (E) FIRE SUPPRESSION ANC DETECTION SYSTEMS (3) OTHER FIRE PROTECTION EQUIPMENT SUD5ET ACTIVITY:
601910C1 FY 1930 SCOPE:
$385K 1.
CONDUCT FULL-SCALE TESTS WITH DETECTION AND EXTINGUISHING SYSTEMS 2.
COMPLETE EVALUATION OF PENETRATION FIRE STOP TEST ME H000 LOGY.
3.
CONTINUE EVALUATION OF GAS SUPPRESSION SYSTEMS AND INITIATE EVALUATION OF WATER SU?PRESSION SYSTEMS.
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SOME SINGLE-CONDUCTOR CABLES " FAILED".
MOST CONNECTOR ASSEMBLIES FAILED; FAILURES MAY BE A (UNCTION OF '
Tile ASSEMBLY, NOT JUST Tile CONNECTOR.
VERY SEVERE MATERIAL DAMAGE OCCURRED FOR CABLES AND CONNECTORS.
NO " MATERIAL" SYllERGIStlS WERE OBSERVED FOR CABLE Atl0 TENSILE SPECIMEllS.
RADIATION IS A PRINCIPLE DAMAGE MECilAlllSM.
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QUAL.lFICATION TESTitiG EVALUATION PROGP.AM TASK 2:
BADI ATION OllALIFICATION S0tlRCE EVALUATION EY_711!EAEIERLDBJECILVES EXTEND SIMULATOR "ADE00ACY" EVAL'UATION TO OlllER CLASS 1E EQUIP COMPLETE / CONTINUE "llEST-ESTIMATE" SIGNATilRE DEFINITION AND CLA EQUIPMENI PESP0f!SE CALCULATIONS lillTIATE SIMllLATOR " TAILORING" STUDIES; DEVISE BENCllMARK CALCULATIONS DEVELOP DOSE AND DOSE-RATE ESTIMATES FOR A GENERIC CONTAINMENT STRUCTURE (SEPARATE FUf!DINd)
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ACCELERATED AGING STUDY FY 79 NEAR-TERM GBJECTIVES CONTINUE / COMPLETE AGING EXPERIMENTS AND MODEl. VERIFICATION ON ELECTRIC CABLE COMPLETE FIRE-RETARDANT AGING, lillTIATE C0ATINGS AGING EVALUATION CONTINl!E ALTERNATE DAMAGE INDICATORS EVALUATIONS EXPAND COMPUTER POLYMER MODEL Al!D APPLICATION EXTEND MEll!0D0 LOGY 10 0 tiler E00lPMENT, 0 tiler ENVIR0fiMEllTS EVALUATE AllERilATE MEll10DS OF AGING EVALUATE NATURALLY-AGED CABLE SAMPLES WilEil AVAILABLE e
00ALIFICAT10N TESTitlG EVALUATION PROGRAM EXPECTED RESULTS IN FY 80 TASK 1:
OllAllFICATION MEll10DOLOGIES ASSESSMENTS Tile COMPLETED FACILITY \\llLL BE USEli T0 INVESTIGATE TYPE-TESTING METil0DOLOGIES AllD SUGGEST TECilNIQUES. METil0DOLOGY TESTING WilL BE EXTEt1DED 10.OlllER CLASS 1 COMPONENTS AS REQUIRED.
INDEPEI1 DENT VERIFICATION TESTS OF SUPPLIERS COMPONENTS MAY BE CONDUCTED AT Tile DIRECTION OF Tile NRC PROGRAM MANAGER.PRELIMINARY EFFORT MAY BE DIRECTED T0\\ LARDS A SEISMIC TEST /METil0D0 LOGY CAPABILITY.
EFFORT Will BE CollTINUED TO DEVELOP AND MAINTAIN Tile TEST FACILITY AS A STATE-0F-IllE-ART 00ALIFICAT10N TEST STANDARDS LABORATORY.
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EXECTED RESULTS IN FY 80 TASK 2:
RADIATION QUALIFICATI0tl SOURCE EVALUATION IllE TilRUST OF Tills EFFORT WILL llE.T0 WARD A MAINTENANCE-0F-CAPABILITY FullCTION 10 PROVIDE TECllNICAL ASSISTANCE FOR SPECIFIC RADIATIGt 00All-FICATION PR0l!LEMS.
TYPICAL OF TilESE MAY I!E "C0tlTINUING" LOCA RADIATION ENVIRONMENTS DEFINITION, GENERIC CONTAltlMENT DOSE / DOSE-RATE ESTIMATES, SIMULATOR TAILORING / DESIGN 1f.o, llENCllMARK CALCULATIONS DEVELOPMENT, AND GENERAL TECllNICAL ASSISTANCE TO NRC.
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EXPECTED RESULTS IN FY 80 IASK_3;_. AECllEI'AIEILAEILGlIllllt Tile SillGLE AND COMllli!ED ENVIRONMENTS TESTING Al[D MODEL V TION lllLL llE COMPLETED FOR ELECTRICAL CAllLE MATERIAL.
Tile EFFORT IllEN WILL llE DIRECTED T0llARD Tile INVESTIGATION OF 0 tiler AGING '
EkVIRONMEllTS, 0 tiler CLASS 1 COMPONENTS / MATERIALS', ALTERNATES TO lilE AGING MODEL (E.G., REQUALIFICATION), AND GENERAL CORRELATION OF Sil0RT-TERM /LONG-TERM TEST TECilN100ES.
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PURPOSE - OBIAlt! DATA UN EXPOSURE AND PERFORMANCE OF SAFELY-IlELATEI) EQUIPMENT IMPLEMF.llTAT10li - IlSE OTE RESEARCll REVIEW Gl100P SPECIFIC WORK ITEMS DETERMlNE EXPOSURE ENVIRONMENT AS A FUNCTION OF LOCA AND TIME DETERMINE PERFORMAllCE OF SAFETY-RELATED EQUIPMENT DUR ACCIDENT DETERMINE E0lllPMENT TO BE REMOVED FOR FilRTilER EXAMINA1 ESTABLISil AND IMPLEMENT A SYSlEMATFE POST-M0pTEM EXAMINATO OF lills E00lPMENT
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WILL llE REQUIRED 10 RESOLVE PRESEflT DIFFICULTIES WITil IMPACT LOCATION.
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o PROVIDES NRC WITil EXPERIE!!CED TECilNICAL EXPERTS FOR RAPID ON-CALL ASSISTANCE, o
ESTABLISilES A DATA LIBRARY OF STANDARDIZED FORMAT AND QUALITY FOR FUTURE ASSESSMENT WORK.
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Il0lSE DIAGl10STICS FOR SAFETY ASSEFSMENT BWR STABILIIY STUDIES OB.lECTIVE:
DElERMINE IF STAlllLITY OF BWRS IS AFFECTED BY CIIANGES It! CORE DESIGN OR OPERATING STRATEGY PURPOSE OF DElERMINE FEASIBILITY OF USil'G NOISE SIGNALS TO ORiil PROGRAM:
DETECT CilANGES Ili CORE STABILITY WITil00T DISTURBING PLANT OPERATION MElll01h OBTAIN DECAY RATIOS OR 0 tiler SIMILAR STABILITY PARAMElERSFROMllEUIRONNOISEiAPRM)SIG11ALSUSlHG TIME SERIES ANALYSIS
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1511 182
Il0lSE DIAGt!0STICS FOR SAFETY ASSESSMENT TliREE-MILE ISLAND.llNIT 2 ASSISTAllCE ORNL TMI-2 PARTICIPAT10li REVEALED SEVEilAL TEClllllCAL lilNDftANCES --
INCOMPLETE UNDERSTANDlHG OF SOURCE (S) 0F TEMPERATURE NOISE, ITS RELI.i!0N TO B0llitlG E BLOCKAGE, ITS DEPENDEllCE ON PPESSURE, FL0ll, ETC.
liiCOMPLETE bliDERSTANDlHG OF PRESSURE fl0lSE Afl0 ITS' DEPENDENCE Dil SYSTEM VOIDS.
INSUFFICIENT EXPERIENCE WITil TilLSE SIGNALS.
LACK 0F BASELINE DATA, PARTICULARLY Ill CFF: NORMAL AND SiluTDOWil STATES.
RESTRICIED SIGNAL ACCESS, LAfK 0F SIGllAL BilFFERING, AliD CONNECTOR ll;COMPAllBILITIES.
LACK 0F MllLII-CllAfNEL TRENDING DEVICE FOR SIGiAL STATISTICAL PROPERTIES.
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Il0lSE DIAGNOSTICS FOR SAFETY ASSESSf1ENTS FY 1980 EXPECTED ACCOMPLISilMENTS PROPOSED t! Elf ACTIVIIIES SilRVEY PERFORMANCE DEFICIEtlCIES OF PRESEN1 PIPE LEAK DETECT 10f! SYSTEMS AND EVALUATr. IMPROVED DETLCTION MElll0DS (PARTICULARLY ACollSTICAL).
EVALU.^!E AND DEfERMINE LIMITAT10flS OF TIME-SERIES SYSTEM ANALYSIS METil0DS, AS APPLIED TO RFACTOR SURVEILLANCE AllD DIAGNOSTICS.
EXTEl,!D PWR BASELillE SIGNATilRE ACQUISIT10t.' ACTIVITIES T0 IllCLllDE TEMPERATURE, PRESSURE AND lil-CORE IlEUIRON (SPND) SIGt!ALS; STUDY SIGt'AL INTER-REl.AT10llSillPS.
EXPLORE Tile FEASIBILITY OF MONITORING FOR CilAtlGES Iti FUEL-TO-COOLA TRANSFL3 llME C0fiSTANT IN A PWR BY MEAllS OF NON-PERTURBING NOISE TEClifil00ES.
PRESENT AN ORNL WORKSil0F ON Tile USE OF fi0lSE TECllN100ES FOR REACTOR SAFETY ASSESSMEllT.
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PWR NEllTRON fl0lSE ANALYSIS GRS (FRG)
PWR TEMPERATURE NOISE HEASilREMEt!TS e
E Cil ( N E Til E R L A fl D S )
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e IllFORMAl. EllROPEAN MEETINGS l
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lillf1All FACTORS RESEARCll SAFETY-RELATED OPERATOR ACT10tlS e
OBJECTIVE ESTAllLISil A DATA BASE WillCil Call BE llSED TO ASSESS PROPOSED CRilERIA OR DEVELOP CRITERIA FOR DETERMINiliG HilElllER OR fl0T REQUIRED SAFETY-RELATED GPERATOR A C T 1 0 flS M llS T B E A ll T 0 t1A T E D.
e APPROACll GAlllIR APPLICABLE DATA FROM OPERATING EXPERIEllCE.
CORRELATE OPERA 10R EXPERIEilCE TO SitillLA10R RESilLlS.
PERFORf1 SiflllLATOR EXPERifiEllIS.
DATA BASE WILL CollSIST OF CALIBRATED SlfillLATOR RESllLTS.
C
" CRITERIA FOR SAFETY RELATED OPERATOR ACTIONS" (ANSI N660, ANS-51,4) il660 CRITERIA ANS COMMITTEE - VENDORS, UTILITIES, A/E, NRC HISTORY 1973 - INITTATED (FIRST DRAFT., STANDARD) 4/74 - AEC " NEGATIVE WITH COMiEliT" VOTE 2/75 - STANDARD REWRITTEN - Ai:S-50 BALLOT CONFENTS
" YEA" Ai:D "NAY" 11/76 - AFTER MANY REVISI0i;S, DRAFT RELEASED FOR TUC 6/77 - COMMITTEE REORGANIZED, NEW CHAIRMAN CURRENT
- REVIEWII;G COMMENTS AND MODIFYING RECUIRED OPERATOR ACTIOi! - MANUAL ACTIONS U9ED TO PREVENT VICLATIOi! 0F DESIGi REQUIREMEi!TS (DESIGN BASIS EVEiiTS, CHAP. 15)
BASIS IS TIME REQUIREM.El!T - EXTENSION OF "TEU-MII;UTE RULE" L0i;GER TIMES FOR - i! ORE SEVERE EVEliTS LESS FREQUENT EVEp!TS LESS FAMILIAR EVENTS I.E., TIME SHOULD INCREASE WITH STRESS AliD DIFFICULTY OF DIAGIOSIS, 1511 190
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e 1511 191
SAFETY-RELATED OPERATOR ACTIONS EVENTS CONSIDERED DIRECTLY APPLICABLE TO N660 EVENT LCT!.0E TIME REQUIREMENTS P.E S/G TUBE RUPTURE
.. INITIATE RCS C00LDOWN AND 9 MINUTES AFTER LOW DEPRESSURIZATION PRESSURE TRIP BORON DILUTIOh IERMINATE MANUALLY s 45 MINUTES LOSS OF A/C POWER PLACE DHRS IN SERVICE (MAY NONE SPECIFIED BE REQUIRED AFTER DIESELS COME ON LINE)
LOSS OF SERVICE RFDUCE POWER AND REMCVE RCP AT LEAST 10 MINUTES WATER FROM SERVICE BWR RELIEF VALVE OPENS TRY TO RECLOSE VALVE; IF NOT ASAP INADVERTENTLY AND SUCCESSFUL SHUTDOWN THE WILL NOT RECLOSE REACTOR INITIATE TORUS COOLING RUPTURE OF PRIMARY SHUTDOWN AND ISOLATE LEAK 1 10 MINUTES INSTRUMENT LINE RUPTURE OF 0FF-GAS CLEAR AREA 0F PERSONNEL, INITIATE BY 1 FINUTE SYSTEM ISOLATE AFFECTED SYSTEM LOSS OF A/C POWER MAINTAIN PRESSURE AND WATER AS REQUIRED LEVELS BY MANUAL OPERATION OF RELIEF VALVES AND RCIC 1511 192
SAFETY-RELATED GPERATOR ACTIONS CHAPTER-15 EVENTS EXAMINED DURING PRELIMINARY ASSESSMENT BWR INCIDENTS PWR INCIDENTS LOAD REJELTION BORON DILUTION ACCIDENT TUREINE TRIP LOSS OF FEEDWATER MSIV CLOSURE LOSS OF ALL A.C. POWER RECIRCULATION PUMP TRIP FIRE IN THE PLANT LOCA LOCA LOSS OF FEEDWATER MAIN STEAM LINE RUPTURE S/G TUBE RUPTURE INSTRUMENT LINE RUPTURE FAILURE OF MAIN CONDENSOR LOSS OF CONTROL ROOM
.0FF-GAS SYSTEM PLANT FIRE OVERPRESSURIZATION LOSS OF CONTROL ROOM INADVERTENT SAFETY INJECTION RELIEF VALVE STUCK OPEN LOSS OF A.C. POWER 9
1511 l93
SAFETY-RELATED OPERATOR ACTIONS WORK ACCOMPLISHED STUDY OF NE60 CRITERIA, THEIR HISTORY AND BACKGROUND.
e IDENTIFICATION OF ACCIDENT EVENTS OF INTEREST AUD SPECIFIC e
OPERATOR ACTIONS.
e SEARCH NSIC DOCKET FILES FOR SPECIFIC EVENTS AT FIVE SELECTED SITES (10 UNITS).
PREPARE OPERATOR SURVEY TO COLLECT QUALITATIVE AND QUANTITATIVE e
DATA (OPINION).
e SITE VISITS - SURVEYS, SITE RECORDS, INTERVIEWS.
e DOCKET SEARCHES, KEY EVENTS, ALL EWR'S AND PWR'S.
e AI.ALYSIS OF DATA (90% COMPLETE).
e LITERATURE SEARCH FOR NON-GUCLEAR DATA.
WORK TO BE DCNE COMPLETE ANALYSIS OF DATA AED LITERATURE SEARCH.
a FREPARE
SUMMARY
OF RESULTS AND CONCLUSIONS OF PHASE I.
e 9
9 1511 194
SAFETY-RELATED OPERATOR ACTIONS
SUMMARY
/ CONCLUSIONS INITIAL DATA e
GENERAL MODEL SEEN AS REAS0fiABLE BY OPERATORS, BUT MAllY (USUALLY VERBAL) RESPONSES INDICATE CONTRIDICTIONS.
e SURVEY RESPONSES SUPPORT IDEA THAT PCTENTIAL SEVERITY OF CONSEQUENCES IS MAJOR FACTOR IN PERCEIVED STRESS.
e SURVEY RESPONSES TNDICATE DIFFICULTY OF DIAGNOSIS DEPENDS ON Af!NUNCIATION.
e OPERATOR PREDICTION OF RESPONSE TIMES IS QUALITATIVELY SIMILAR TO LIMITED PERFORMANCE DATA COMPILED TO DATE -
DIFFUSE LOG-fl0RMAL DISTRIBUTION.
e OPERAT0k PREDICTION OF RESP 0i;SE TIMES IS QUANTITATIVELY MORE OPTIMISTIC THAN PERFORMANCE DATA - EUT rat:GE IS SIMILAR.
e WE HAVE ONE EXAMPLE (INADVERTENT SAFETY INJECTION) 0F AN EVEliT FOR WHICH WE HAVE BEEN ABLE T0:
(A)
DEFIt!E A SPECIFIC ACTION TO BE FEASURED IN AN EVENT OF REASONABLY DIRECT APPLICABILITY TO N660; (B)
COLLECT A SIGNIFICANT NUMBER OF DATA POINTS.
THE RESULTS SUGGEST A LOG-NORMAL DISTRIEUTION t.'ITH A RANGE THAT EXTENDS TO TIME VALUE COMPARABLE TO THE N660 CRITERIA (95% = 7.4 MINUTES).
1511 195
SAFETY PELATED OPERATOR ACTI0f1S CCI!CLUSIONS AND RECOMMEt!DATIONS ON tis 60 CRITERIA C0tiCLUSIONS e
THE CONCEPTUAL MODEL OF OPERATOR RESPONSE IS REASONABLE FOR SOME EVE!!TS, INAPPLICABLE FOR OTHERS.
e THE USE OF A TIME MARGIli FOR OPERATOR RESP 0i!SE IS A REASONABLE APPROACH FOR INTERIM CRITERIA, I.E.,
UliTIL A THOROUGH HUMAN FACTORS STUDY IS COMPLETED.
e ItiCREASING THE TIME MARGIN WITH STRESS AMD DIFFICULTY OF DIAGi10 SIS IS REASONABLE; QUAMTIFICATION OF STRESS AliD DIFFICJLTY IS EXTREMELY DIFFICULT.
OPERATOR IUPUT IS VALUABLE.
e ORhL DOES NOT HAVE EN0 UGH DATA TO PAKE A JUDGEMENT
~
ON THE CURREliTLY RECOMMENDED Tlf.E MARGINS.
- HCWEVER, EATA TO DATE DO NOT SUPPORT SUGGEST10i! 0F TIMES SIGNIFICANTLY GREATER THAN CURREi!TLY PROPOSED (E.G., OkE-HOUR).
RECOMMENDATIONS FROCEED WITH DEVELOFMElli 0F INTERIM CRITERIA BASED ON BEST AVAILABLE INFORMATION.
I'0DIFY TIME MARGINS AS DATA FROM THIS AND/0R CTHER PROGRAMS ACCUMULATES, e
INVESTIGATE POSSIBLE WAYS TO ACCOMMODATE
" SYMPTOMATIC RESP 0fiSE".
15))
196
SAFETY-REl.ATED OPERATOR ACTI0flS C0flCLUS10!!S Oli DA1 A AVAILABILITY EXPENSIVE CCNCLUSI0li _ DATA COLLECTION IS NOT IMPOSSIBLE,. lust VERY fitfffEUL-T
_fRDIlLBLS POTENTIALFOR" PURE"DATABASEISSMALL-JUDGEMENT,EXTRAPOLATIONNbCESSARY.
2 DEFINIfiG EVENTS / ACTIONS - t!0 " STANDARD" EVEl:TS OR PROCEDURES.
e DEFINiflG WilAT 10 MEASllRE - JUDGEMENT REQUIRED; CASE BY CASE VARIATION.
e DOCKET SEARCillNG ll0N-TRIVIAL - NO SINGLE S0llRCE; LOGGING PROBLEMS.
e CURRENT LER FORM llAS LESS DATA 111AN OLDER FORMS OR SPECIAL REPORTS.
e INTERPRETA110!10F SITE DAlA, ESPECIALLY COMPUTER OUT.PUT. REQUIRES e
SKILLED SITE PERS0 TINEL.
e SITE ACCESS AND AVAILABIL11Y OF SITE PERS0l!flEL LIMITED.
e OPERATOR SURVEY BY SITE PERSONNEL, NOT EFFECTIVE.
SITE RECORDS KEEPII;G PLANS GOOD; EXECllTI0ff ONLY FAIR.
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s IlUMAN FACTORS RESEARCll lillMAN ENGINEERING AND SYSTEM PERFORMANCE OBJECTIVES ASSESSMENT OF ACCIDENT INVESTIGATION TECilll10llES i
DEVEl0PMENT OF PROGRAMS FOR TRAll!!NG llRC INSPECTORS AtlD LICEN9EE REVIEWERS IN Tile PRINCIPLES AND PROBLEMS OF ERGON 0 METRICS EVALUATION OF lillMAN ERRORS e
Lz w
IlUMAN ENGillEERiflG AND SYSTEM PERFORMAllCE Tile PRINCIPAL ELEMENTS ARE:
~
IluMAN ERRORS IN MAINTENANCE IlUMAN ENGINEERING AND SYSTEM PERFORMANCE IDENTIFY PROBLEMS AllD ESTABLISil PRIORITIES FOR WORKING lillMAfl PERFORMANCE RELIABILITY PROBLEMS ALARMS:
IlUMAN RESPONSE RELIABILITY (1980)
STUDY ALARM SYSTEMS AND TilEIR llUMAN FACTORS IMPLIEAT10iis I
c=
FY 1980 PROGRAM ASSUMPTIONS PROGRAM: WRSR TITLE:
FI'; NO:
AE119 HUMAN FACTORS METHODS DEVELOPMENT CONTRACTOR:
EGL5 I..::
etkTd: I65HO Si NRC TECHNICAL MONITOR:
W. S. FARMER PRINCIDAL INVESTIGATOR:
CATHERIt:E STEGR-v 119 :.
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J.
EUJ3ET ACTIVITY:
60191001 Fi 1950 PROGRAY ASSUMPTIONS:
590K 1.
COMPLETE THE ANALYSIS OF FIELD COLLECTED CATA FOR HuAN RELIAEILITY IN MAINTEN!';CE AN3 CALIBRATION ACTIVITIES AT ODERATING NUCLEAR POWER STATIONS AND PROVIDE A SUMMADY REPORT IDE!.TIF.YING THE iH-nC..,,..7 a.
It.,
. S:
R,.,,. I,- 1. Y O H.,.... P. r 0.,.,.n,,i. -
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REVIEW AENOR',iAL OCCURRENC.E REPORTS, LECs.AND CD'.'PLIANCE REPORTS TO y n N.. I Y n.. :.na-..... - H,,,.,... P -- -,.... -. L.... I i Y I.e.
L n,
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an n:
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3.
REV I E'a-THE CURRENT PRACTICE AND USE OF LIGHTS, ALAo.MS A'.O ANNUNCI ATORS oc
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CACILITATE HU.AN' MACHINE INTERACTION AND MINIMIZE E:00.S.
DEVELOP RECO'"'ENDATIONS AND CARRY OUT LAEORATORY OR FIELO Ex?E IMENTS TO TEST HYo0 THESES.
IDENTIFY MEANS FOR IM.: ROVING THE HL"A:-MACHIt.E INTERACTION IN NUCLEA:. CONTROL R00"S.
A.
PROVICE AN ANNUAL REPORT FOR THE PROGRAM.
1511 201 s
e FY 1980 PROGRAM ASSUMPTIONS PROGRAM: WRSR T'^TLr:
HUMAN ENGINEERING REVIEW FIN NO:
A1194 GROUP CONSULTATION CONTRACTOR:
SAND:A
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SITE: ALBUQUEROUE STATE: NEW MEXICO NRC TECHNICAL M0ii: TOR:W. 5. FARMER PR:NC PAL INVESTIGATOR:
A. D. SWAIN OBJECT:VE: ASSIST NRC IN HUMAN ENGINEERING RESEARCH REVIEW BUO3ET ACTIVITY:
60191001 510K Ty 1980 ROGRAM ASSUMPTIONS:
PARTICIPATE AS A CONSULTA';T IN DISCUSSI$NS AT MEETINGS OF THE REACTOD.
1.
SAFETY RESEARCH DIVISION'S HUMAN ENGINEER:NG REVIEW GROUP.
2.
ASS:ST RSP BY PROVIC:N3 AN ONGDING REVIEW OF !M 0RTANT WOPK IN THE ARE: or HU'te ENGINEERING AND SUM'iARIZING TH:5 REV!EW EITHER AT HUtiAN EN3INEERING REY!EW GROUP MEETINGS OR BY LETTER REPORTS.
3.
REVIEW ENGLISH AESTRACTS OF FOREIGN LANGUAGE ARTICLES RECE!VED BY NRC UNDER INTERNATIONAL EXCHANGE J3REEMENTS AND RECO'tMENCING WHETHER THE A:.TICLE SHOULO BE TRANSLAIED.
A.
DRJVID!NG PR033R">ATIC ASSISTANCE TO RSR BY OEVELOPING RECO WENDATIONS AND CEF:N:N3 RE00:REMENTS FOR HU' TAN EN3:NEERIN3 RESEARCH.
O I511 ?D
EOLL0lflNE
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STUDY OF Tile PROCESS OPERATOR -
BY J. RASMllSSEN (DENMARK)
OPERATORS - SYSTEM AND TASK ANALYSIS -
BY P. BLOMBERG (SWEDEN)
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\\,c RESEARCll SUPPORT BRAilCll C0ilTRACTOR OPERATitlG FiltiDS (t000)
OP_FRATIONAL SAFETY E
RECOMMENDED FY 1979 FY 1980 FY 198]
FIRE PROTECTI0li fili 5 385 890 1f455 QUALIFICATION TESTIllG 780 600 IlUMAN FACTORS 215 195 800 NOISE DIAGNOSTICS 190 170 600 O
250 2000 VALVE STUDIES COMPUTER CONTROLS 0
0 355 ALTERNATE CONTROL SYSTEt1S 0
0 1150 IE SUPPORT 0
0 250 OPERATIONAL SAFETY STl1 DIES 0
O 1050 ALTERNATE Sil0TD0HN SYSlEMS 0
0 900 DIAGNOSTIC SYSTEMS 0
0 550 G
Z LEClifilCAL SUPPORI 715 585 1500 TOTAL OPERATit;6:
2,3fl5 2,185 10,800 3
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