ML19269D031
| ML19269D031 | |
| Person / Time | |
|---|---|
| Issue date: | 11/01/1978 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1595, NUDOCS 7902260603 | |
| Download: ML19269D031 (19) | |
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,!'!I f SUBC0fEITTEE MEETING WASHINGTON, D. C.
.NOVEIGER 1,1978 A meeting of the ACRS Subcommittee on Reactor Operations was held in Washington, D. C. at 1717 H Street, N. W. on November 1, 1978. The purpose of the meeting was to review the Office of Nuclear Regulatory Research programs being conducted by the Research Support Branch, and to hear a briefing by the Divisich of Operating Reactors. Notice of the meeting appeared in the Federal Register, Volume 43, No. 201, October 17, 1978. The schedule for discussion and a list of attendees at the meeting are attached to the minutes. No written statements were received from members of the public, and there were no requests from members of the public to make oral statements. The subcomittee did not issue, approve, or receive any reports during this meeting. A copy of the vu-graphs shown during the meeting is attached to the office copy of these minutes.
MEETING WITH THE NRC RESEARCH SUPPORT BRANCH STAFF (OPEN) (8:30 a.m. - 3:1'5 p.m.)
Mr. Gary Bennett, Research Support Branch Chief made a short introductory presentation on the organization of the Research Support Branch and its activities.
The Research Support Branch is one of five branches under the Assistant Director for Water Reactor Safety Research. Research Support Branch functions cover four areas: Operational Safety Research, a program in 3D flow distribution, Technical Support Services, and c.ertain management support functions. The Operational Safety Research Program is being conducted in four areas: fire protection, qualification testing evaluation, human factors and noise diagnos-tics. Mr. Mathis suggested that the Research Support Branch also follow work being perfor:nd on noise detection of water leaks.
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QUALIFICATION TESTING EVALUATION ( Transcript pgs. 10-44)
Objective's of the qualification testing evaluation program are to obtain data to improve the technical basis for standards and Reg. Guides for Class 1 E electrical equipment and to provide a data base for future changes, as appro-priate.
The Qualification Testing Evaluation Program is composed of three tasks:
qualification methodology assessment, radiation qualificatinn sonrce evaluation, and an aging methodology model study.
The qualification methodology assessment program deals with the question of se"quential versus simultaneous exposur6 of components or mater 1ars to environments that might be expected during a loss of coolant accident.
The IEEE 323 Standard allows sequential exposure to environmental conditions.
An alternate way of qu 'ification testing'of the class lE equipment is to perform a simultaneous exposur<. test in which. accelerated thermal and radiation aging are combined followed bf combined radiation, steam, and chemical spray exposure.
Tests conducted at Sandia were designed for both sequential and simultaneous test-ing.using typical equipment and the IEEE 323 specifications. Arrhenius techni-ques.were used for accelerated aging. Thus far 9 tests have been completed, five of which are significant. The tests have been performed on cables, cable connector assemblies, cable splices and insulation. Some work was performed on paints, coatings, and lubricants.
Important conclusions thus far from the synergism testing are as follows:
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for types A and B cable and types R and S connector assemblies, no functional synergisms were observed from the available data. In the case of the connec-tor assemblies, certain data would seem to indicate that the sequential test profile is a little worse.than the simultaneous test. There were no function-al synergisms observed for the cable splice assemblies. Some single connector cable failures were observed in the sequential tests. The type S & T cable connectors consistently failed during the LOCA simulation testt, but the failures may be partially due to the way in which they were assembled.
An important part of the program at Sandia is to identify failure criteria
since it has an important effect on related tests and equipment. The criteria _
used in the tests is based on an elongation test because embrittlement of the Z insulation materials may be an indicator of electrical shorting. Other con-clusions reached are that aging can affect the LOCA simulation test results, radiation is a principal damage mechanism, and synergistics effects do not appear significant in LOCA type tests of electrical cable.
The existing test facility has certain limitations and Sandia is in the process of upgrading it. The new facility will have new radiation sources and source positioning, and will be able to accomodate larger and more diverse class 1 E components. It will also allow selectable radiation dose rates.
A program is also underwt.y to put together a data package which includes a description of the different kinds of. safety equipment that are installed in plants. This would then be used for future evaluation testing of methodology.
EPRI has a related program in this area to perform a similar study on BWR's.
Objectives for FY 79 are listed on figure 1.
Mr. Bennett next discussed the radiation qualification source evaluation program also being conducted by Sandia. The objective of this program is to determine how well the existing radiation similators simulate radiation effect. This is tied in with the overall qualification testing of equipment and the radiation dose which has to be applied for aging and LOCA tests. The technique that Sandia is using is to calculate the LOCA release rates, compute the radiation spectra and then test and correlate the calculations. This yields depth dose profiles which can then be compared with the similators. Conclusions reached from the energy release rate calculations are: effect of length of irradiation prior to a LOCA is not significant on energy release rate less than one day after a LOCA, length of_ exposure can be significar.t after 1 day, and effects of fuel,
fission product, transmutation, and fissile isotope depletion are not signi-ficant. Also, the method of treatment of the daughter product is important.
Preliminary conclusions by Sandia are that greater precision is needed in the
. specification of input parameters. The data base on nuclide tractionation should be improved and the significance of beta radiation should be assessed.
For FY 79, it is planned to extend the similator adequacy evaluation to other class lE equipment and do further work on the best estimate signature definition
4 and class 1E component response calculations. There may also be an effort to develop dose rate estimates for a generic containment structure.
The 3rd task of the qualification testing' work is concerned with an aging methodology model. The objective of this task is to determine a reasonable way to simulate 20,30 and 40 year aging on components and to develop a method-ology which can be used by NRR to assess applicant submittals.
Work is underway to develop a single environmental aging test covering radiation, temperature, and humidity. Some combined environmental aging testing including rate effects and some way of evaluating aging in addition to testing elongation will also be done.
Some of the early results from Sandia show that elongation decreases smoothly and monotonically towards zero. They also noted that materials can be become quite brittle as they approach zero elongation, and that humidity doesn't seem to have any effect. Finally, mechanical degradation seems to depend on the integrated dose. Mr. Bennett emphasized that the program is aimed at developing a test not testing hardware.
Thus far, low level radiation testing and aging facilities have been completed.
Aging experiments and model verification have been performed for typical electrical cable insulations and jackets.
In FY 79, the plan is to continue the aging experiments, investigate aging effect on fire retardant coatings and continue work on finding other means of determining indication of damage besides tensile testing. Also, work will begin to observe other equipment and other environments as well as continuation of the evaluation of alternate methods of aging.
In response to a question from Mr. Mathis, Mr. Bennett said that the qualification testing evaluation program budget was $775,000 in FY 78, and $78L.000 is planned in FY 79.
FIRE PT4fECTION RESEARCH (transcript pas 44-97)
Mr. Bennett reviewed the Fire Protection Research Program. The three principal objectives of the program are to "obtain data in support of Regulatory Guides and standards for Fire Protection in LWR power plants, to establish an improved technical basis for modifying guides and standards, and to obtain data to aid licensing decisions. The Fire Protection Program is largely an outgrowth of the report on the Browns Ferry fire and includes the eight subelements shown on figure 2.
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The programs being conducted at Sandia are addressed to evaluating the adequacy.
ofcable.trayspacinginReg. Guide 1.75toevaluatevariationsinseparation1 distance; to evaluate electrical initiated fires in stacked trays in conduits, and evaluate sinole tray and stacked tray exposure fires. Also inchded are investi-gations of fire retardant coatings and. fire suppression system studies.
Mr. Bender asked if there were. an/ plans to conduct tests of detectors.
Mr. Bennett replied that the Office of Standards Development is going to conduct a program on detectors.
Mr. Bennett reviewed the series of tests conducted in 1976 to evaluate the Reg. Guide 1.75 separation distance in terms of protecting redundant safety divisions from an electrically initiated fire. As a result of these tests it was found that the Reg. Guide 1.75 separation between redundant divisions is adequ.ce to prevent tray to tray propagation with electrically initiated cable fires. Mr. Bender comented that he is interested in electrically initiated fires in things like large motors and connectors. Mr. Bennett replied that in future years tests on equipment other than cables are planned. Mr.
Bender also suggested that it would be useful to consolidate at one place all information on fire testing.
Mr. Bennett next described the tests run in July, 1977 using the stacked tray configuration with Reg. Guide 1.75 separation between the redundant cable tray divisions with an exposure fire initiation. Two 70,000 BTU per hour propane burners were located beneath the bottom tray. An asbestos fire barrier was placed between the bottom tray and the next tray above so that the burners could not propagate the fire to any of the trays above the ignition tray.
After five minutes, the two burners were turned off and the insulating barrier was removed. About 10 minutes after the fire started in the ignition tray the tray above caught fire, and about 1 minute after that, the fire in the ignition tray went out. The fire proceeded up through the stacked cable trays to the redundant safety division in about 25 minutes. About 35 minutes after the test began, the structure holding the cable trays collapsed and the test was concluded.
It is believed that the upper two trays caught fire before the structure collapsed. Mr. Bennett said that for the electrically initi. ed fires, the tests confirmed the NRC Staff position in Reg. Guide 1.75. He said that the test also confirmed the NRC Staff position on exposure fires, because cre-
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dit has not been allowed solely for separation.
Mr. Bennett then showed a 16 minute film showing portions of the electrically initiated and exposure cable fire tests and some of the cable coating tests.
Next Mr. Bennett described the. tests performed on fire retardant coatings.
Some small scale tests were run at first and later full scale tests were run at Sandia, using r sentially the same cable tray configurations as used in the earlier exposure tests. The chief difference was that in the fire retardant coatings test, the burners were cycled 5 minutes on and 5 minutes off for a maximum of 6 cycles. The full scale tests involved single and two tray config-uration with fires initiated by the propane burners and diesel fuel (simulating spilled fuel). Six different coatings were applied according to the manufacturers' directions and tested. Both 383 qualified and non-383 qualified cables were used.
Some tests were also run with no fire resistant coating applied. Although there was propagation observed during these tests, all the fire retardant coatings offer some resistance to combustion when compared with uncoated cable.
Mr. Bennett also described a series of fire shield tests performed as part of the saW activity. Various combinations of a ceramic wool blanket placed over covered and uncovered ladder trays using 383 qualified and non-383 qualified cables in single and two tray configurations were tested.
As in the case of the coatings, all the barriers tested offer some protection.
No propagation was observed with the non-383 cable although there was ignition.
In trays with the 383 qualified cable there was no ignition.
Mr. Bennett desctibed the new facility Sandia is developing which will have the capability for testing water and gas suppression systems.
The first test is scheduled for May, 1979. The objectives of this program, which involve both vertical and horizontal cable trays, are to obtain information on the effectiveness.of water ano other extinguishing agents and to investigate potential damage effects on safety-related equipment.
'The work is to be divided into two sub-tasks: the first of which is to con-struct the test bed, and the second is to collect information on fire suppression systems and reconnend a test program.
. DEVELOPMENTANDVERIFICATIONOFFIRETESTSFORVERTICALCA2LESYSTEMS
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_( transcript pgs 98 - 137)
Mr. W. Christian of Underwriter's Laboratories discussed the work Underwriter's Laboratories is doing involving vertical configurations and flame spread tests. The research is being conducted in four areas:
1.
burner sensitivity experiments; 2.
full cable tray experiments; 3. flame spread tests using an IEEE 383 type configuration, and 4.
proof tests of vertical cable tray systems to verify fire protection measurements for protecting vertical cable trays.
The first two areas of research were designed to revise the standard IEEE 383 test to provide a more realistic test result. The effort was focused on attempting to develop a test using full cable trays not the single layer of cables as is the current IEEE 383 test requiren.ent. Eventually, two 35,000 BTU per hour burners with a fuel air ratio of 8 to 1 were decided upon.
Using this burner arrangement, a single vertical tray, 40% filled with cables experiments were run to try to produce reproducible results for IEEE 383 type tests. This effort us not successful, and it was decided not to use a full tray configuration for IEEE 383 type qualificaticn tests. Instead, it was decided to improve the 383 test procedure to make it a well defined standard screening test.
A serles of tests were performed varying different parameters and using three different cable types. After additional work, UL intends to make reconnendations on how to standardize the equipment and procedures so that the IEEE 383 test will be more reproducible.
Mr. Christian discussed the large scale proof test for a vertical cable system with fire barriers and fixed automatic fire detection suppression system that UL conducted as part of their program in support of the Sandia research.
In this test, 5 vertical cabla trays were placed in the corner of a 15 foot hiah room; the trays were fully loaded with IEEE 383 qualified and some IEEE 383 non-qualified cables, and each tray.vas completely wrapped with a 1-inch ceramic wool blanket. One ionization type and one photoelectric type detector were in-stalled on the ceiling 15 feet from the corner. There were also three open head sprinklers in the ceiling. At each sprinkler location, there were three tell-tale sprinklers. The main sprinklers were to be manually actuated when
8-all three tell-tale sprinklers actuated. A spill fire was simulated using
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2 gallons of liquid heptane in a 25 square foot pan placed at the base of the cable trays. 90 seconds after the start of the test, the heptane was essentially gone from the pan, but flaming continued for about 40 minutes around the base of the trays. Mr. Christian indicated that the heptane was probably soaked up by the ceramic wool blanket. The detectors actuated 11 and 14 seconds after the fire began. Two of the tell-tale sprinklers in one loca-tion actuated just under a minute after the fire started. No other sprinklers in the array actuated at any time, so no water was applied to the fire.
After the test was concluded, the sprinklers were tested to make sure that they were not defective. The highest temperature measured inside the cable tray array was 141 degrees F, but according to photographs taken the temperature did become hotter in some areas. Two conductors shorted together in tray number 3, and there were some erratic measurements in tray number 1.
Mr.
Christian believes that the erratic measurements were the result of some faulty connectors in the wiring above the ceiling of the trays. Megohm tests showed all cable to cable resistances in the hundreds of thousands of ohms range except for one conductor in tray number 5 which was 100,000 ohms. Mr.
Christian indicated that the test was not conclusive as far as insulation break-down. Visual examination of the cables revealed insulation damage in 4 of the 5 trays. However, except for the short that occurred in tray 3 there was no indication of electrical trouble. Tray 5, in which there was no visible damage, was separated from the other 4 trays.
Mr. Christian showed a short film of the vertical cable tray tests. He con-cluded by saying that UL has proven that the ceramic wool blanket provioes adequate thermal insulation from a spill type fire for the vertical tray con-figuration tested. However, there is a problem of the fire getting underneath the ceramic barrier in the area of contact with the floor.
3 Further tests are being deferred until the results of this first test are studied.
NOISE DIAGNOSTICS ( transcript pg. 138-156 Mr. R. Kryter discussed the program in noise diagnostics that is supported by the Reactor Safety Branch. He said that although financial support
9 comes from the Reactor Safety Branch, most of the work is oriented to 7
problems that originate with other NRC groups. The development group of the Instrumentation and Controls Division at Oak Ridge provides consulting services in noise diagnostics to the NRC Staff to assist them in evaluating data and proposals generated by vendors.and licensees. Specifically, the assistance is provided as on-call technical assistance through the Division of Operatirig Reactors and the Division of Systems Safety, confirmatory research, and standards development.
Mr. Kryter reviewed some of the major accomplishments of the development group prior to FY 1978 and accomplishments for DDR, the Division of Engineering Standards, and the Division of Reactor Safety Research in FY 78.
Mr. Kryter snid that in FY 79, they anticipate continued work investigating the Fort St. Vrain temperature and flux oscillations. Fundamental studies of loose parts monitoring systems and some more work on BWR stability and sto-castic code development is expected to continue. As an outgrowth of work performed last year, they anticipate further work in neutron noise signature acquisition for PWR's and its transfer from one plant to another.
HUMAN FACTORS RESEARCH ( transcript pg. 158-173)
Mr. W. Farmer, NRC Staff, reviewed the two programs that are being started in Human Factors Research. He discussed some of the concerns that have risen in the last few years, noting that the human engineering review group was established in 1976. The review group identified three areas where the focus of NRC work should be placed. The three areas are to reduce potential for human errors, to establish a data base for human reliability risk assessment, and to provide a technology base for developing guides and standards..Mr. Farmer briefly reviewed some programs sponsored by other groups in the nuclear industry an't other NRC Staff divisions.
The two human factors research programs being sponsored by the Reactor Safety Branch are being performed at EG&G and at Oak Ridge National Laboratory.
The study at EG&G has thus far focused on identifying training programs that might be useful to I&E to equip their staff with a better understanding of human factors.
In FY 79, EG&G is going to perform a study to review human factors involved in maintenance errors. In FY 1980, plans call for a correla-
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tion study between compliance and human error in plant operation, and a study =
on the relationship between alarm systems and human response.
The Oak Ridge Study is concerned with safety-related operator actions. The coal is to evaluate human perfomance under strass in sarious safety occurrences and to determine what time limits would be reasonable to expect operators to take prompt and sufficient action.
Other possible human factor s_udies in FY 80, for which there are no specific plans are: operator fatigue criteria, advanced control room designs which involve a new field of machine-human interaction, and improved ways that the operator can conduct diagnostics.
MODEL5 OF HORIZONTAL CABLE TRAYS EXPOSED TO FIRE PLUME (transcript pc. 174-193)
Mr. L. Hunter of the Applied Physics Laboratory discussed the work that he has been doing in conjunction with Sandia and Underwriter's Labs. The study performed by Mr. Hunter is an effort to model and interpret the exper cental results from Sandia and Underwriter's Laborato-ies. Using this, it is hoped to aid in the design of the experiments and make occasional pretest predictions.
The models developed by Dr. Hunter predict the delay time to the onset of flamable gas evolution and the gasification rate attainable when direct ignition does not occur. The models apply to horizontal insulated cables and cable trays as well as coated cables and trays. Calculations were made from traditional theory of heat transfer and the results are presented in the form of charts. Calculation formulas for the quantity of flamable gas evolved from an exposed tray when that tray does not directly ignite have also been developed. The analysis in-cludes a qualitative explanstion of the behavior of the fire ball in horizontal trays.
Ignition,which is a complicated phenomenon resulting from an interraction between aerodynamics and chemical reactions.is not predicted by.Mr. Hunter's models.
The charts for determining delay time have been developed for the one and three cable studies at Sandia. Curves showing the qualitative effect of the overall cable rad,ius and the longitudinal conduction in the cable were also illustrated.
Mr. Hunter next discussed the delay times calculated with the charts for the cables tested by Sandia. He noted that the calculated delay time to the onset of flamable gas evolution is well within the lifetime of the plume.
. management.
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He reviewed the source of the many licensing actions performed by DDR and gave some typical examples of issues requiring significant manpower.
He also described some typical license amendments, both NRC proposals and licensee proposals on which D0R acted.
Mr. Stello next discussed the feedback of operating experience into licensing.
In addition to informal comunications within NRR and elsewhere in the NRC Staff, the DOR established a fomal feedback system in 1976.
The syster '.onsists of operating experience memoranda informdng NRR of potential'iy significant findings from operating reactors and general information memoranda which provide general information obtained from operating reactors for senior NRR management.
Mr. Stello discussed the systematic evaluation program, its objectives, the plants involved, and typical items on the list of topics being reviewed.
He also reviewed the approach taken with resoect to making decisions as to what needs to be backfitted. Members of the subcommittee indicated that it is not clear what operating experiences are used in the SEP.
Mr. Stello reviewed activities associated with core reload reviews.
He mentioned that the most difficult question to deal with in tems of reloads is when the fuel vendor is changed.
He also mentioned that some vendors are submitting complete reload topical reports.
Next, Mr. Stello discussed backfitting. There is a fomal process in 10CFR 50.109, for backfitting, but this has been very seldom imposed. Within NRC the Regulatory Requirements Review Comittee makes decisions on backfitting.
It looks at changing requirements and, depending on the situation, makes a determination as to which reactors need backfitting. Backfitting also goes on during the CP and OL reviews.-
The Division of Operating Reactors is spending a large amount of manpower on inservice ins,pection and testing. Plants are required to conform with Section 11 of the ASME code. Because of various situations, licensees ask for, relief from the requirement. Currently, an average of about 15 inservice relief re-quests and about 25 testing review requests are granted per plant. Mr. Stello
said that he is attempting to obtain a change in rule to permit the inservice inspection and inservice testing to be updated at 120 month intervals rather E than the current 40 months for ISI and 20 months for IST.
Mr. Stello discussed the Division of Operating Reactors response to incidents.
He discussed some typical examples such as the Bl.W weld wire, the BWR channel box vibration problem, and the BWR pipe crack problem in 1975. Finally,
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Mr. Stello reviewed the conditions and the reviews performed for plants which request an increase above licensed power.
If the power level is higher than the design power level, then all the engineered safety features and the containment response are viewed; if the increase in power t..,es not go beyond the design power, then all of the transients in Chapter 15 are re-analyzed. In response to a suggestion from Mr. Bender, Hr. Stello agreed to provide a written list of all the specific things that are re-analyzed when a power increase is requested.
Mr. Haas briefed the Committee on the activities of the Quality Assurance Branch which is part of the Division of Project Management.
Responsibilities of the QA Branch are to develop specific criteria for inclusion in Standard Review Plan, Section 17.2 for operational quality assurance. These criteria are used for judging the acceptability of operational quality assurance.
The applicant provides a description of what he proposes to implement and this is reviewed by the Quality Assurance Branch.
Implementation cf the requirements is inspected by the Office of Inspection and Enforcement. Normally, there is an on-site and off-site quality assurance organization for each plant. The Quality Assurance Branch reviews the application to assure that proper QA con-trols have been established for plant operation maintenance, modifications re-fueling etc.
With respect to pre-operational testi g and inspections, it is also a two stage review. The more extensive review is performed at the OL stage, where the specific tests to be conducted are reviewed as well as staffing levels and qualifications of people. This is covered in Standard Review Plan Section
,14.2 and several Reg. Guides. The Office of Inspection and Enforcement inspects what the applicant is doing and provides information regarding the successful completion of pre-operational tests.
The plant operational organizations are reviewed at the CP stage and again ha ~ i + a-i a
=-a actah14 chart in Raa. Guide 1.8.
at the OL stage.
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El Specific requirements for plant review and audit function are contained in Normally, there is an off-site and on-site review organization.
There are specific qualification requirements for personnel who perform the review and audit function.
In response to a question from Mr. Bender, Mr. Haas said that the QA program applies to design changes, but quality assurance in this situation would be done by an organization other than the plant operations quality assurance organization.
Details of the discussions during the meeting can be found in the transcript of the meeting. The transcript is a stenographic, uncorrected record of the discussions at the meeting, and no responsibility is accepted for errors or inaccuracies of statements or data contained in the transcript. Copies of the transcript are available at the NRC Public Document Room and can be purchased from ACE Federal Reporters, Inc., 444 North Capitol Street, Washington, D. C.
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LIST OF ATTENDEES ACRS H. Etherington, Chairman W. Mathis M. Bender R. Wright, DFE NRC H. L. Ornstein P. R. Matthews R. L. Ferguson E. D. Sylvester W. S. Farmer T. R. Quay R. E. Salomon A. S. Hintze T. Murley W. Haas OTHERS W. Christian, Underwriters Labs R. Kryter, ORNL R. Luna, Sandia L. Hunter, APL a
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REFERENCES l.
45 Vu-graphs used by RSB to illustrate fire protection research.
2.
5 Vu-graphs used to illustrate RSB organization.
3.
64 Vu-graphs used by RSB to illustrate qualification testing evaluation.
4.
9 Vu-graphs.used by RSB to illustrate human factors.
5.
5 Vu-graphs used by RSB to illustrate other activities.
6.
9 Vu-graphs and summary of "Modals of Horizontal Cables and Cable trays exposed to a Fire Plume," by L. W. Hunter, Applied Physics Laboratory.
17.
12 vu-graphs used by Instrumentation and Controls Division, ORNL to illustrate current activities.
8.
31 Vu-graphs used by UL to illustrate vertical cable tray fire tests.
9.
35 mm slides used by UL to illustrate vertical cable tray fire tests.
10.
16 minute film clip of fire protection research at Sandia.
- 11. 6 minute film clip of vertical cable tray fire tests at UL.
- 12. 24 Vu-graphs used by DDR to illustrate current D0R activities.
- 13. 6 Vu-graphs used by DPM to illustrate cuality assurance and plant operations review activities.
- 14. Draft Program Plan for the Research Support Branch.
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W SANDIA QTE PROGRAM QUALIFICATION METHODOLOGIES ASSESSMENTS (CONT.)
FY 79 NEAR-TERM OBJECTIVES COMPLETE TEST FACILITY UPGRADE, MAKE OPERATIONAL COMPLETE / DOCUMENT COMMISSION-REQUESTED CONNECTOR TESTS PROVIDE SPECIFIC NRC-REQUESTED CONFIRMATORY TESTING COMPLETE SHORT-AilD LONG-RANGE TEST PLANS, PROJECT DESCRIPTIONS CONDUCT INITIAL METHODOLOGY TESTS PERFORM CLASS 1E EQUIPMENT SENSITIVITY EVALUATION ACQUIRE OFF-SITE DATA; CONDUCT OFF-SITE TESTING DEVELOP REQUALIFICATION TESTS FOR NATURALLY-AGED EQUIPMENT D
3 EVALUATE " STATISTICAL" OUALIFICATION METil0DS h
W W
W RESEARCH SUPPORT PRA'!CH FIRE PROTECTION RESEARCH PROGRAM SU9 ELEMENTS e
OBTAIN DATA ON EFFECTIVENESS OF CA9LE-TRAY SEPARATION M,'
CRITERIA e
OBTAIN DATA ON EFFECTIVENESS OF
- CONDUITS, FIRE
- RARRIERS, AND PENETRATION FIRESTOPS e
OBTAIN DATA ON EFFECTIVENESS OF C0ATING MATERIALS e
OBTAlli DATA ON FIRE RETARDANCY OF AGED MATERIALS g( e OBTAIN DATA ON IEEE STD 383-1974 AND DEVELOPP.ENT OF IMPROVEP SMALL-SCALE CABLE-SYSTEM OUALIFICATION TESTS e
OBTAIN DATA ON EFFEC 1fVENESS OF SAFETY-RELATED EQUIPMENT (OTHER THAN CABLE)
WilEN SUBJECTED TO EXPOSilRE FIRES
'e OBTAIN DATA ON FIRE DETECTION SYSTEM PERFORMANCF.
Ne OBTAIN DATA ON EFFECTIVENESS OF WATER AND OTHER FIRE-d EXTINGUISHING AGENTS
W W..
i RESEARCH SUPPORT BRANCH LONG RANGE PLANNING OPERATIONAL SAFETY RESEARCH s
SUBELEMENT FY-1980 FY-1981 FY-1982 FIRE FIRE STOPS
, OTHER SAFETY EQUIPMENT
. FULL SCALE TESTING EXTINGUISHING SYSTEMS
. DETECTORS / EXTINGUISHING (AGING EFFECTS)
OTE COMPLETE LOCA
. HON-LOCA
. CONFIRM NON-LOCA NOISE ANALYTICAL METHODOLOGY
, INITIAL STABILITY MODEL
. SUPPORT TO NRR/SD DIAGNOSTICS, LAB TESTS HUMAN ASSESS DIAGNOSTICS
. SIMULATOR STUDIES
. AUTOMATION STUDIES FACTORS VALVE TEST PLAN
. CONSTRUCTION
. TESTING DESIGN STUDIES Il