ML20003F923
| ML20003F923 | |
| Person / Time | |
|---|---|
| Issue date: | 08/08/1980 |
| From: | Budnitz R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RIL-096, RIL-96, NUDOCS 8104230893 | |
| Download: ML20003F923 (16) | |
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SW MEMORANDUM FOR: Harold R. Centon, Director Office of Nuclear Reactor Regulation FRCM:
Robert J. Budnitz, Director Office of Nuclear Regulatory Research RESEARCH INFOR.vaTICN LETTER #ec - ACECUACY OF CURRENTLY (USJECT:
UTILIZED FADIATION TEST SOURCE 5 TO SIMULATE THE LOSS-f CF-COCLANT DESIGN BASIS ACCICEST INTROCUCTICN This reecrandur. transmits a sun.ary of the results of a ecmpleted pcrtien of ttle tdC Cualificatien Testing E !uaticn (QTE) Program relating ::
the ade:;uacy of currently utilized radiatien simulators to cenduct i
radiatien qualificatien of safety-related c::uipment.
Padiation qualification is one part of the loss-of-coolant accident I
(LOCA) design basis accident qualification required by NRC Regulatior.s.
Synergistic effects between radiation and other accident environments I
However, for those may not allcw the separation of accident conditions. cases where se upon viiich the adequacy of radiation simulators can be judged.
Simulator adequacy can be judged by comparison of the radiation magnitude, rate, spectra and particle type of the calculated and simulated source, terms. It is also possible to show equivalence of the source terms by showing that similar damage to materials would result from exposure to both source terms. The latter approach is utilized in the research covered by this RIL.
The research includes the development of a calculational method for determining the radiation magni.tude, spectra, and particle type as a function of time that would result from the release assunptions defined in Regulatory Guide 1.89.
In addition, scoping radiation dose rate calculations have been made for a typical empty containment structure.
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. Harold R. Denton Also, a study was performed to establish typical depth-dose profiles for a polymeric material in the configuration of a typical class IE cable insulation, and to identify the credible radiation damage mechanisms.
Finally, an assessment was made of the differences in the currently used test source characteristics and Regulatory Guide 1.89 radiation character-istics to identify damage mechanisns and to evaluate radiation simulator adequacy when utfitzed for LOCA qualification testing, including the related radiation aging.
SUMMARY
The research results support tna following conclusions:
The calculations of radiation magnitude, spectra, and particle type as a function of time that result from the hypothetical r 1.
in Table I.
The calculated total dose is at least a f actor of two higher than
- However, the dose currently utilized for LOCA qualification testing.
2s the total integrated calculated dose is approximately equal to the total dose resulting from currently utilized fuel melt m LOCA.
The calculated dose consists of a significant beta radiation component while cobalt-60, which is the typically utilized radiation s 3.
The calculated peak dose rate from combined sources is a factor of This results from the instantaneous release assum 4.
Guide 1.89.
From a material damage point of view, as might occur on a typical polymeric material such as a cable insulation mat higher calculated dose rate should not yield significantly more damage for any given dose level.
However, the effects on electrical system performance due to the i
generation of induced electrical noise, caused by j
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i For example, LOCA qualification accurate duplication of the circuit.
testing of cable is usually conducted with a low resistive load which would not show the effects of noise pulses.
The damage to safety-related materials resulting from a predominately j
beta-radiation source should not be any greater than the damage that 5.
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would result from gamma radiation because the mecha i
by either of the following:
The total gamma radiation dose can be established at least at the level of the total calculated dose, which is approximately a.
6 45CMR for a 30-day exposure, if the Regulatory Guide 1.89 release assumptions are used, or l,
The relative damage effects of gamma and beta radiation for the A comparison of the b.
specific cor ponent can be established.
damage effects on a typical polymeric material, such as electrical cable insulation, showed the effects to be similar on a pe
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most of the damage on the surface, as opposed to a more unifon-e volume basis.
i damage that results from a cobalt-60 exposure, was not evaluated.
With the exception of electrical cable, most instru 4
the effects of beta radiation, but may result in secondary For radiation (Bremsstrahlung) which should be account i
be shown to be conservative.
i The following are the areas for which additional research is underway 6.
or planned:
- a. ' Best-estimate calculations are being perfomed to establish a more realistic LOCA qualification source tenn.
Tests will be conducted to compare the damage caused by bata and gamma radiation on generic cable designs.
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of research will be an evaluation of the secondary (Bremsstrahlu radiation effects on instruments with protective covers that would be protected from the dire,ct beta radiation, but not the secondary l
x-rays.
LOCA tests may be conducted with materials a t
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Sasis for the Research_
The specific research upon which this RiL is based is outlined in the Research Support Branch Plan (Enclosure 1) for Qualificatica Testing The overall QTE program is based on the research need19,1977), the identified in the research request, (5077-7, dated August Evaluation.
NRR program sup: ort letter fron E. G. Case, dated January 5,1978, as C user groups.
well as through censultation and forr.a1 review with the NR f
The QTE Review Group has been, and continues to be, the focal point f b th fomulation of the research program and evaluation program result that fem the basis of this RIL.
The specific user requirement for the research ccnducted to date on th rantee that 3dequicy of radiation simulators is based on the need to gua l
current loss-of-coolant :::f dent qualification testing confoms with the Regulat:ry Guide require-ents and that if censee qualification programs l
are technically souno.
The calcu!ation Of the radiation field inside containment fo11cwing.a LOCA is directly itnked to the assumed release from the core and th I
specific contaircent geometry.
fission product release from a design basis acciden acceptable fission product source term.
The concern that equipment be not only designed to the appropriate accident conditions, but also be qualified under these conditions, i
culminated eventually in the development of IEEE 323-74 as the ac industry standard. The standard was written with this position as a f accident radsforPWR' sand 2.6x10jacourse-o-t d d suggeste exposure of 1.5 x 10gx A to the s an ar rads for BWR's, but focal point. Append Part no specific bases were provided for these values in the standard.
of thir standard was adopted in Regulatory Guide 1.89, " Qualifica Class '.E Equipment for Nuclear Power Plant " which was issued in No However, the radiation source term contained in Appendix A was-spe_ifically excluded from adoption in Regulatory Guide 1.89 b l
1974 did not confom to a TID-type of fission product release.
The current NRC staff pcsition is that cobalt-60 or similar simulator, d
tl which is the most widely used radiation test source, can a equa e yThe ca I
simulate the LOCA radiation' environment.
as part of this research and the damage assessment perfomed technical basis upon which the adequacy of simulators can be jud 1
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made along with calculations to detemine the sensitivity of these measurernents to reactor fuel composition, operating duration, power I
level, and treatment of progeny, and are documented in Reference 1.
Calculations of absolute magnitudes and rates require that certain The containment structure was plant-specific assumptions be made. assumed to be an empty cylinder (
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baving an inside radius of 17.7m, inside height of 63.6m, and concrete
- i walls that were 1.lm thick. This results in a total free volume of 0cm3 (2.2 million ft ).
The calculations were made for a 3
power level of 40C0 MW (themal) and were based on the Regulatory Guide 6.25 x 10 1.89 assumptien that the accident resulted in an instantaneous and unifom release, the effects of engineered safety features were ignored, and containment leakage was not accounted for.
i The prijnary code used for the fission product energy release calculations RIBD calculates isotopic concentrations resulting from fission sources with nomal down-chain decay by beta-emission and isomeric was RIBD-II.
i The transfers and interchain coupling resulting from n-gama reactions.
.pmgram library used is based on the ENDF/B-IV fission product data set and contains 818 fission product isotopes for each of fou For each isotope, cross-sections for both fast and themal systems.
l only the average gama and beta energy release is given.
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The CINDER code was used for calculation of the composite spectra of the Using the concentrations of all fission product fission products.
nuclides of interest fmm the RIBD code, the spectra were calculated c
using CINDER and the spectra data that are included in the ENDF/B-IV The method of calculation of the gama and beta spectrum has l,
library.
been designated as the GABAS spectrum code.
Using the Regulatory Guide assumptfor.3 and the codes discussed above, the calculations were made for the accident case where the airborne t.I source 13 unifonnly distributed throughout the containment volume, the late-out source is unifonnly' distributed on the containment wall, and he waterborne source is unifomly distributed in a pool on the containment 1
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. Harold R. Denton Using these assumptions and those explicitly defined in the Regulatory Guide, the spatial dependence of the maximum radiation was floor.
calculated and occurs on the containment center line near t water pool.
codes and the regulatory guide assumptions to verify the results.
Material Damaqe Calculations In order to determine the adequacy of currently utilized radiation simulators, quantitative values of depth-dose and charged particle distribution were detennined.
The The calculations of depth vs. dose were made using the code SANDYL.
code computes the photon and electron transport and energy deposition within the system defined by use of a Monte Carlo computational method.
The depth-dose calculations were made for a typical organic material exposed to both the calculated LOCA accident source release and typical 1
The material chosen for the depth-dose and radiation test simulators.
damage evaluetion was a polymeric material in the configuration of a typical class IE electrical cable consisting of a co:per conductor surrounded by an ethylene-propylene rubber insulator and a i
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in the cable, was determined for the spectral extremes of the calculated jacket.
Although the LOCA radiatien release and for Cobalt-60 and Cesium-137.
calculations were made for a specific cable, they are typical of the results that would be obtained for any exposed organic materials.
i Damace Assessment e
The final determination of the adequacy of a radiation simulator must be f
based on a comparison of the resulting damage and damage mechanisms to The the exposed material from both the simulators and acci This was t.ccomplished by a review of the literature damage mechanisms.on radiati.on effects on materiais and by evaluation of the p l
effects resulting from radiation particle interactions.
l The damage assessment was based on the depth-dose a However, calculations discussed.
environment was at typical reactor ambient conditions.
usi.ng typical LOCA conditions of 143*C and 60 psia showed that the The temperature and pressure did not significantly effect the results.
following speci.fic damage mechanisms were examined:
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Ionization occurs by either collisions of beta radiation source.
The particles and electrons or by photon-electron interactio amount of beta radiation while the electrons resulting from the simulator sources are generated primarily by photon-electron interactions The beta radiation deposition is primarily in the exposed material.
in a region near the material surface while the photon-electmn interactions are more unifomly distributed within the exposed material.
The effect of the radiation-induced nonunifom charge distribution First, the unbalanced charge could cause the following problems.
distribution could cause noise in electrical circuitry; and second, the nonuniform charge distribution if sufficiently large could i
cause the breakdown of the material dielectric strength.
Since there is a potential difference with regard to charge buildup distribution between the hypochetical source-term and radiation simulators usi,ng only gama radiatien, a quantitative assessment of this damage mechanism was made. Estimates of charge distribution were made as a function of raterial thickness for extremes of LOCA spectra profiles and for Cobalt-60.
strength of a typical polymeric cable material was then determined by calculating the resultant electric field caused by the charge huildup.
Specific The question of radiation induced signals was addressed These calculations were radiation in typical polymeric material.
ll made using an empirical relationship for conductivity change with Leakage dose rate that is generally true for polymeric materials.
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current was estimated by using the change in conductivity and lf calculated estimates of the electric field caused by charge buildup.
l1 The discharge of the excess charge that can buildup as a result of the nonunifann radiation field can cause noise pulses which could l
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An estimate of this effect the operability of the equipment.
effect was made by assuming that the discharge occurred between the l
insulation and copper conductor of a typical plant signal cable.
l There is a possibility that the extent of damage caused by the 1.ncrease in temperature,within the material resulting from the l
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radiation source attenuation is dependent on the type of t
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The then the resulting temperature profiles wiU be different.
e hypothetical LOCA source tenn, based on the accident assumptions in Regulatory Guide 1.89, results in a temperature profile that is
. higher at the surface and lower within the material than would be Because of this the case with a Cobalt-60 radiation simulator.
difference, a quantitative assessment of the effects of unequal e'
temperature distribution and of ter:perature rise es included in this research.
There is a possibility that the extent of damage caused by bulk
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energy deposition and the spatial energy deposition within the material is dependent on the energy spectra and on the type of g
The hypothetical LOCA radiation source, based on I
radiation source.
the Regulatory Guide 1.89 assumptions, is a time-varying energy I
spectra and is different from the energy spectra of radiation Because of this difference, a quantitative assessment simulators.
of the relationship of damage to energy spectra was included in I
In general, radiation energy deposition in poly ers tnis research.
can result in the femaden of ions, the breaking or creaticn of The creation of ions molecular chains, and the evolution of gas.
The and the resulting charge buildup are included in item 1.
remaining radiation induced effects are included in this category, The calculated source tem includes a sizeable beta contributfor f
that is generally not considered in equipment qualification prograns and this contribution, in tenns of total dose, is higher than the total dose currently utilized in radiation qualification programs.
The calculated gamma radiation, however, is lower than that currently used in radiation qualfication prograns.
Table I shows that the calculated peak dose rates for both gama and beta radiation are higher than the gama radiation dose rate The currently utilized in radiation qualification programs.
l calculated source tenn is characterized by changing gama and beta J
spectra with the hardest spectra occurring about a minute after the release and the softest spectrum occurring in about 4 days, and the emissions are, in general, different from the line spectra radiation source simulators currently utilized in radiation qualification l
programs.
There is a possibility that the extent of damage caused by the gradient of the energy deposition is dependent on the type of 4.
Material degradation can result from stress l
radiation source.
generated by differential material shrinkage caused by nonunifonn energy deposition and may be enchanced by a loss of elasticity
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also caused by the radiation. Nonunifom energy deposition results in greater energy deposition close to the surface than in the interior of the cable.
If there is insufficient cable elasticity to allow for the shrinkage, cable damage can occur. Thus, in a typical polymeric electric cable, nonunifonn shrinkage could result in circumferential or radial cracking and possible material damage.
Experimental data were evaluated to establish typical elasticity and shrinkage loss rates using several commonly used safety related electric cables. Elongation was used as a measure of elasticity.
t RESULTS Calculated Radiation Scurce Tem l
The objective of the research was to calculate a definitive radiation scurce tem based on tne NRC Regulatory Guide release assumptions, and to cetemine the ade:;uacy of currently used radiation qualification Considerable data i
simulators to duplicate this radiation environment.
were generated with regard to the source tem definition.
Table I represents a cumulative 30-day value of gauna and beta dose and The data in the references give the maximum values of dose rates.
detailed time histories separated as waterborne, airborne, and plate-out i
sources and also shows their spatial dependence. The generally accepted i
source tem, which is used for LOCA qualification, is also included in j
Table I.
First, what Examination of Table I raises the following two questions.
is the sensitivity of the calculated source tenn values to the assumptions that were made to obtain these values; and second, what is the significance of the differences between the calculated and simulated ' source tem values.
Source Tem Calculation Sensitivity The references include a significant amount of parametric calculations showing the sensitivity of the results to changes in release assump-n Examination of these data lead to the tions and core parameters.
rates and conclusions that the total energy released, the releasr5 spectra are not significantly changed by fuel compositten, power level, 3
However, the total t
duration of operation, and treatment of progeny.
I energy released, the release rates, and spectra are significantly changed g
by the nuclide fraction release assumptions.
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l Damace Assessment Examination of Table I indicates significant numerical differences in the calculated source tem and the one currently utilized by industry.
The significance of these differences was studied to detemine if the i
source tem, currently utilized in radiation qualification programs, f
duplicates the dar. age that would be caused by the calculated source Certain damage mechanisms that could be dependent on the choice j
of radiation source were studied to detemine if the magnitude of the tem.
g differences is significant.
Charce Builduo Effects The charge buildup is primarily the result of a nonunifom electron radiation dcse deposited in the cable. This cccurs because of the The specific mecnanism is radiation attenuation within the material.
that the secordary emission leaving a unit volume is not balan'ced by I
Table I shows that tne secondary emissien from an adjacent unit volume.
calculated LOCA source is comprised of a sizeable beta radiation centribution therefore, this effect would be more prevalent with the calculated LOCA source because the beta radiation is attenuated more than the gamma radiation from a simulator source such as Cobalt-60.
(Although Cobalt-i 60 does decay by beta emission, the beta particle energy is low, and j
therefore, remains within the cobalt material or encapsulation.)
For the hypothesized LOCA and Cobalt-60 radiation sources to be equivalent.
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either the damage caused by charge buildup will have to be shown to be negligible or the charge distribution will have to be shown to be the l
same.
The problem in making thh comparison is complicated because the hypothesized LOCA spectra varies with time and as a result, the charge distribution is also changing. The time histories of the beta and gamma spectra, calculated as part of this research, show that the spectral extremes of the radiation energy occur at approximately 1 minute and 4 days.
Beta and gamma energies were calculated for these spectral extremes for typical polymeric cable material. Using experimental and analytical
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data from other work referenced in the radiation effects literature, charge buildup distributions were' calculated for both radiation sources.
From these charge buildup distributions, values of electric field were
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'l The effect of the charge buildup on the signal integrity of a cable used in a safety related system was not detemined because it is so strongly For example, charge buildup in a dependent on the type of circuit.
cable which is part of a high impedance circuit could result in a serious However, since current practice is to noise problem or signal error.
not duplicate the circuit or even the circuit impedance, this effect should not be a factor in evaluating currently utilized radiation sources.
This issue may be further evaluated as part of the OTE program at a later date.
Tem:erature Effects L
The radiant energy from the calculated LOCA source and a simulator, such E
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. as Cobalt-60, is deposited within the exposed materiaAs described in tne previc material temperature.
prcfiles of tne calculated and Ccbalt-60 sources are not the sa e, and therefore, the resulting increase in temperature or temperature gradient i
of the exposed material may be different. For example, the calculated i
LCG source rasults in nore energy deposited closer to the surf ace of As part of tne research, the raterial than d:es the Ccbalt-60 source.
these tec;erature effects were analyzed for both the calculatec and For the maxieur. dose rate calculated, the Ccbalt-60 radiation sources.
effects due to temperature gradients caused by both the nonunifem i
energy deposition and the themal lag in the transfer of heat to the surrounding environment are not significant enough to cause material i
damage.
Total Enerev Effects Bulk energy deposition within the test specimen is dependent on th energy spectra.and Cobalt-60 radiation sources, including variations, as a function of depth in a typical polymeric cable material,
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For the calculated LOCA source, which has a time varying energy spectra, These energy deposition the maximum energy deposition rate was used.
t values were used to calculate the heat flow across a generic electrical These data show that cable and to determine the resulting temperatures.
if poor conduction to the surrounding envirorcent is assumed, damaging cable insulation temperatures could be reached and that the cable temperature l
would depend primarily on the heat transfer coefficients between the j
cable and the envirornent, and to.a much lesser degree on the energyThus, the que deposition profile or the deposition rate.of cable due to energy deposition l
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consideration.
include the heat contribution f-un any direct steam impingement on the cable. The cabined effects of high radiation dose rate and energy deposition from LOCA steam could cause damage to exposed polymeric materials.
In any event, the effect is not strongly dependent on the I
radiation source as long as.the energy deposition rates are equal.
An additional damage mechanism' that was considered for a typical polymeric material is the total chemical and mechanical change such as scission It was shown that these changes were not significantly and crosslinking.
i dependent on the type of radiation source as long as the total energy This is because the energy interaction mechanism i
deposition is similar.
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in polymeric material is the same for both radiation sources.
Mechanical Stress Effects I
One additional damage mechanism that was considered for a typical polymeric material is the mechanical stress that could result frcm J
This could result in surface cracking and loss of insulating qualities The amount of shrinkage is dependent when subjected to a LOCA environment.
on the energy deposition profile as a function of depth below the surface, and hence, the damage mechanism, if important, could be source dependent.
l Experimental data giving shrinkage and loss of elasticity of certain commonly used cable insulating materials were obtained showing that for Also these materials the shrinkage is small (generally less than 5%).
I it was noted that most of the shrinkage occurred early in the experiment Thus, although before the material lost any appreciable elongation.
insulating materials may become brittle (as evidence by a loss in elongation) as a result of radiation, there should be no significant differential Since
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stress caused by shrinkage and accompanying loss of elongation.
these data do not include all currently utilized insulating materials, the issue should be reconsidered when additional data become available.
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EVALUATION AND RECOMMENDATIONS The data obtained in this RIL and an advanced copy of the RIL have been reviewed with members of the Qualification Testing Evaluation Review e
Group.
The data discussed in this RIL are of primary concern for the review of LOCA qualification for safety-rela.ted equipment that must operate following I
the accident for periods of time that result in a high accumulated dose.
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The data show that the use of Cobalt-60 or other similar simulator as a radiation source for LOCA qualification should be accepted by the licensing staff. There is no basis, at this time, for requiring a change to a different radiation source. However, the question of whether or not some additional beta radiation c:alification should be required from the licensee for exposed polymerit. :naterials cannot be fully answered without further work, which is in the current QTE research Requiring additional beta radiation qualfication, pending program.
these results, is not warranted since current estimates show that even I
with surface damage on exposed polymeric cable, the cable probably will f
be able to perfom its function because the bulk elasticity probably A series of scoping tests is planned in
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will not degrade significantly.
the near future to verify this supposition.
The scurce tem calculations show that the Regulatory Guide 1.89 release assumptions result in a source comparable to current core melt releases.
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This is an overly conservative requirement, particularly for equipment The core melt designed to teminate the LOCA and prevent a core melt.
source tem should be reserved for instruments that may be required to perfom following an untenninated LOCA. Also, the instantaneous release assumptions result in dose rates that are difficult to achieve during equipment qualification testing. Research has been conducted on obtain-ing "best estimate" source tem values. A report on this work will soon be available and could be used in specifying a multi-level qualification source term.
NUREG-0588 " Interim Staff Position on Environmental Qualifications of Safety Related Electrical Equipment," published for coninent in Decembdr l.
1979, proposes a radiation source tem different than that derived from The differences arise from i
Regulatory Guide 1.89 by the Sandia work.
the assumptions made with regard to the deposition of the source and the allowance for the beneficial effects of emergency containment spray.
The NUREG-0588 source term is somewhat less conserva11ve than the Sandia 1
If either the NUREG-0588 or Sandia source tem calculated source tenn.
is used for licensing, it should be modified as quickly as possible with a best estimate source, including reasonable conservatism.
s a ton Current LOCA qualification practice does not include be given to addressing
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I the question of system perfonnance degradation that may be caused by adiation induced electrical noise. The current research program on tt b t est ate rc t c
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by reducing the maximum dose rate.
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COORDINATION CONTACT For c6 ordination of further evaluation of these results and for discussion and future experiments, the, reader is advised to contact Mr. Ronald Feit, g
Qualification Testing Evaluation Research Program Manager, telephone number (301) 427-4272.
A e h
Robert J. Budnitz, re I
Office of Nuclear Regulatory Research I
Enclosure:
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Current Practice
- Theoretical Calculations *
Gama 150-200 MR 50 MR 400 MR Beta 2.
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Gamma
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Beta 3.
Spectra Usually a fixed Time dependent Energy energy source Spectra
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These data are based on a 30-day exposure.
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References:
i 1.
L. L. Bonzon, Sandia Laboratories, " Radiation Signature Following the Hypothesized LOCA, SAND 760740," NUREG/766521, October 1977.
l 2.
L. L. Bonzon, William Buckele-:, Sandia Laboratories, " Evaluation of Simulator Adequacy for the Radiation Qualification of Safety Related Equipment, SAND 791787," NUREG/CR 1184, June 1980.
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