Regulatory Guide 1.77
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| Issue date: | 05/31/1974 |
| From: | Office of Nuclear Regulatory Research |
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May 1974 U.S. ATOMIC, ENERGY COMMISSION
REGULATORY GUIDE
DIRECTORATE OF REGULATORY STANDARDS
REGULATORY GUIDE 1.77 ASSUMPTIONS USED FOR EVALUATING A CONTROL ROD
EJECTION ACCIDENT FOR PRESSURIZED WATER REACTORS
A. INTRODUCTION
reactor is normally limited by the-design of the control rod system to a value well below that which would result Section 50.34, 'Contents of applications: technical in serious damage to the reactor system. However, a information," of 10 CFR Part 50, "Licensing of Pro postulated failure of the control rod system provides the duction and Utilization Facilities," requires that each potential for a relatively high rate of reactivity insertion application for a construction permit or operating which, if large enough, could cause a prompt power license provide an analysis and evaluation of the design burst. For U0 2 fuel, a large fraction of this generated and performance of structures, systems, and components nuclear energy is stored momentarily in the fuel and of the facility with the objective of assessing the then released to the rest of the system. If the fuel energy potential risk to public health and safety resulting from densities were high enough, there would exist the operation of the facility. General Design Criterion 28, potential for prompt rupture of fuel pins and the
"Reactivity Limits," of Appendix A, "General Design consequent rapid heat transfer to the water from finely Criteria for Nuclear Power Plants," to 10 CFR Part 50, dispersed molten U0 2 . Prompt fuel element rupture is requires the reactivity control system to be designed defined herein as a rapid increase in internal fuel rod with appropriate limits on the potential amount and rate pressure due to extensive fuel melting, followed by rapid of reactivity increase to assure that the effects of fragmentation and dispersal of fuel cladding into the postulated reactivity accidents can neither result in coolant. This is accompanied by the conversion of damage to the reactor coolant pressure boundary greater nuclear energy, deposited as overpower heat in the fuel than limited local yielding nor sufficiently disturb the and in the coolant, to mechanical energy which, in core, its support structures, or other reactor pressure sufficient quantity, could conceivably disarrange the vessel internals to impair significantly the capability to reactor core or breach the primary system.
cool the core. General Design Criterion 28 also requires that these postulated reactivity accidents include con The Regulatory staff has reviewed the available sideration of the rod ejection accident unless such an experimental information concerning fuel failure thresh accident is prevented by positive means. olds. In general, failure consequences for U0 2 have been insignificant below 300 cal/g for both irradiated and This guide identifies acceptable analytical methods unirradiated fuel rods. Therefore, a calculated radial and assumptions that may be used in evaluating the average energy density of 280 cal/g at any axial fuel consequences of a rod ejection accident in uranium location in any fuel rod as a result of a postulated rod oxide-fueled pressurized water reactors (PWRs). In some ejection accident provides a conservative maximum limit cases, unusual site characteristics, plant design features, to ensure that core damage will be minimal and that or other factors may require different assumptions both short-term andlong-term core cooling capability which will be considered on an individual basis. The will not be impaired.
Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position. For the postulated control rod ejection accident, a
B. DISCUSSION
mechanical failure of a control rod mechanism housing is assumed such that the reactor coolant system pressure The rate at which reactivity can be inserted into the would eject the control rod and drive shaft to the fully core of a uranium oxide-fueled water-cooled power withdrawn position.
USAEC REGULATORY GUIDES Copies of published guides may be obtained by request Indicating the divisions desired to the US. Atomic Energy Commission. Washington, D.C. 20545.
Regulatory Guides we issued to describe and' make available to the public Attention: Director of Regulatory Standards. Comments and suggestions for Improvements In thes guldes are encouraged and should be sent to the Secretary methods acceptale to the AEC Regulatory staff of implemnenting specfifc parts of the Commission's regulations, to delineate techniques used -by the staff In of the Commission. US. Atomic Energy Commission, Washington, D.C. 20645 evaluating specific problems or postulated accidents, or to provide guidance to Attention. Chief, Public Proceedings Staff.
applicants. Regulatory Guides are not substitutes for regulations and compliance with them Isnot required. Methods and solutions different from those set out In The guides era issued in the following ten broad divisions:
the guides will be acceptable If they provide a basis for the findings requisite to 1. Power Rleactors S. Products the Issuance or continuance of a permit or license by the Commission. 2. Research and Test Reactors
7. Transportation
3. Fuels and Materials Facilities
8. Occupational Health
4. Environmental and Siting 9. Antitrust Review Published guides will be revised periodically, asappropriate, to accommodate L. Materials Ond Plant Protection 10. General comments and to rflect new Information or experience.
A sufficient number of initial reactor states to consequences of this accident for a pressurized water completely bracket all possible operational conditions of reactor.
interest should be analyzed to assure examination of upper bounds on ultimate damage. In areas of uncer
C. REGULATORY POSITION
tainty, the appropriate minimum or maximum para meters relative to nominal or expected values should be Acceptable assumptions and evaluation models for used to assure a conservative evaluation. The initial analyzing a rod. ejection accident in PWRs are presented reactor states should include consideration of at least the in Appendices A (Physics and Thermal-Hydraulics) and following: B (Radiological Assumptions) of this guide. By use of these appendices, it should be shown that:
Zero power (hot standby) - Beginning of Life (BOL)
and End of Life (EOL); 1. Reactivity excursions will not result in a radial Low power - BOL and EOL; average fuel enthalpy greater than 280 cal/g at any axial Full.power - BOL and EOL. location in any fuel rod.
The effects of the loss of primary system integrity 2. Maximum reactor pressure during any portion of as a result of the failed control rod housing should be the assumed transient will be less than the value that will included in the analysis. It should- also be shown that cause-stresses to-exceed the Emergency Condition-stress failure of one control rod housing will not lead to failure limits as defined in Section III of the ASME Boiler and of other control rod housings. Pressure Vessel Code.'
The approach that should be used in the radiological 3. Offsite dose consequences will be well within the analysis of a control rod ejection accident is to deter guidelines of 10 CFR Part 100, "Reactor Site Criteria."
mine the amount of each gaseous radionuclide released to the primary containment and, with this information 1 Copies may be -obtained from the American Society of in conjunction with the procedures set forth in Mechanical Engineers, United Engineering Center, 345 East 47th Appendix B of this guide, to determine the radiological Street, New York, New York 10017.
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APPENDIX A
PHYSICS AND THERMAL-HYDRAULICS
the magnitude of the transient. Pressure and temperature The assumptions described below should be applied are mainly significant with respect to their effect on the in evaluating the physics and thermal-hydraulic behavior amount of reactivity inserted if there exists a positive of the reactor system for a control rod ejection accident.
moderator coefficient.
I. The ejected rod worth should be calculated based
5. ' The fuel thermal properties such'"as fuel-clad gap on the maximum worth rod resulting from the following heat transfer coefficient and fuel thermal conductivity conditions: (a) all control banks at positions corres
.should be conservatively chosen, depending upon the ponding to values for maximum allowable bank inser transient phenomenon being investigated. For conditions tions at a given power level and (b) additional fully or of a zero or positive moderator coefficient (usually at partially inserted misaligned or inoperable rod or rods if beginning of life), for example, high heat transfer allowed by operating procedures. Sufficient parametric parameters would reduce the Doppler feedback and studies should be performed to determine the worth of increase any positive moderator feedback effects and the most reactive control rod in each rod group for hence tend to increase the magnitude of the reactivity different control rod configurations, both expected and transient. For a negative moderator coefficient, high unexpected. The worth of single rods in rod groups heat transfer parameters could cause the magnitude of should be evaluated during startup physics tests and the transient to decrease if a given quantity of heat compared with values used in the rod ejection analysis. produces more feedback in the moderator than in the The accident should be reanalyzed if the rod worths fuel. In the consideration of pressure pulses which may used in the initial analysis are found to be noncon be generated, high moderator heating rates could cause servative. Calculated rod worths should be increased, if significant pressure gradients to develop in the moder necessary, to account for calculational uncertainties in ator channels. In computing the average enthalpy of the parameters such as neutron cross sections and power hottest fuel pellet during the excursion for power cases, asymmetries due to xenon oscillations. low heat transfer would be conservative.
2. The reactivity insertion rate due to an ejected rod 6. The specific heat of U0 2 has been determined should be determined from differential control rod experimentally and is a deterministic factor in the worth curves and calculated transient rod position versus calculated amount of stored energy (enthalpy) in the time curves. If differential rod worth curves are not fuel. Recommended values in the range of 25 to 902fC
available for the reactor state of interest, conservatism are the data reported by Moore and Kelly (Ref. 2). In should be included in the calculation of reactivity the range of 900 to 2842!C, the data obtained by Hein insertion through consideration of the nonlinearity in and Flagella (Ref. 3), Leibowitz, Mishler, and Chasanov reactivity addition as the rod passes through the active (Ref. 4), and Chasanov (Ref. 5) are recommended for core. The rate of ejection should be calculated based on the heat capacity of the fuel. These recommended values the maximum pressure differential and the weight and' are for clean core conditions. Possible variation in the cross-sectional area of the control rod and drive shaft, specific heat due to bumup should be investigated and assuming no pressure barrier restriction.
appropriate values used, if necessary.
3. The calculation of effective delayed neutron frac
7. The moderator reactivity coefficients due to voids, tion (Beff) and prompt neutron lifetime (2*) should be coolant pressure changes, and coolant temperature based on the well-known definitions resulting from per changes should be calculated based on the various turbation theory, such as those described by Henry (Ref. assumed conditions of the fuel and moderator using
1), using available experimental delayed neutron data standard transport and diffusion theory codes. If no and averaging by the fraction of fission in the various three-dimensional space-time kinetics calculation is per fissionable materials. In cases where the accident is quite formed, the reactivity feedback due to these coefficients sensitive to Peff (where the ejected rod worth >Peff), the should be conservatively weighted to account for the minimum calculated value for the given reactor state variation in their spatial Importance in the missing should be used. For smaller transients, conservatism in 'dimension(s). If boric acid shim is used in the moder the value should include consideration of not only the ator, the highest boron concentration corresponding to initial power rise (which increases with decreasing P), but the initial reactor state should be assumed.
also the power reduction after the trip. Similar don siderations should also be applied to determine an appropriately conservative value of 2*to be used. 8. The Doppler coefficient should be calculated based on the effective resonance integrals and should include
4. The initial reactor coolant 1pressure, core inlet corrections for pin shadowing (Dancoff correction).
Calculations of the Doppler coefficient of reactivity temperature, and flow rate used in the analysis should be should be based on and should compare conservatively conservatively chosen with respect to their influence on
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with available experimental data such as those of Hell. characteristics and changes in flux shapes should be strand (Ref. 6). Since the Doppler coefficient reflects the investigated, and the conservatism of the flux shapes change in reactivity as a function of fuel temperature, used for reactivity input and feedback, peak energy uncertainties in predicting fuel temperatures at different deposition, total energy, and gross heat transfer to the power levels should be reflected by conservatism in the coolant should be evaluated. Also, sensitivity studies on applied value of the Doppler coefficient. If no three variations of the Doppler effect, power distribution, fuel dimensional space-time kinetics calculation is performed, element heat transfer parameters, and other relevant the reactivity effect of spatially weighting the core parameters should be included.
average temperature rise in both the axial and radial directions should be calculated. 13. The pressure surge should be calculated on the basis of conventional heat transfer from the fuel,. a conserva
9. Control rod reactivity insertion during trip versus tive metal-water reaction threshold, and prompt heat time should be obtained by combining the differential generation in the coolant to determine the variation of rod worth curve with a rod velocity curve based on heat flux with time and the volume surge. The volume maximum design limit values for scram insertion times. surge should then be used in the calculation of the If the rod worth curve (reactivity vs. depth of insertion) pressure transient, taking into account fluid transport in is not obtained from a "true" representation (i.e., an x, the system, heat transfer to the steam generators, and y, z, t or an r, z, t calculation); the conservatism of the the action of the pressurizer relief and.safety valves. No approximate calculation should be shown. The dif credit should be taken for the possible pressure reduc ference in the depth of insertion at zero power and at tion caused by the assumed failure of the control rod full power should be accounted for in calculating the pressure housing.
available scram reactivity.
14. The number of fuel rods experiencing clad failure.
10. The reactor trip delay time, or the amount of time should be calculated and used to obtain the amount of which elapses between the instant the sensed parameter contained fission product inventory released to the (e.g., pressure or neutron flux) reaches the level for reactor coolant system. It should be assumed that clad which protective action is required and the onset of failure occurs if the heat flux equals or exceeds the value negative reactivity insertion, should be based on maxi corresponding to the onset of the transition from mum values of the following: (a) time required for nucleate to film boiling (DNB), or for other appropriate instrument channel to produce a signal, (b) time for the causes.
trip breaker to open, (c) time for the coil to release the K.,
rods, and (d) time required before scram rods enter the core if the tips lie above the core-reflector interface. The margin to DNB is expressed in terms of a departure from nucleate boiling ratio (DNBR). The
11. The computer code used for calculating the tran DNBR at any position in the hottest channel is the ratio sient should be a coupled thermal, hydrodynamic, and of the DNB heat flux to the actual heat flu
x. The DNB
nuclear model with the following capabilities: (a) incor heat flux should be evaluated using correlations based on poration of all major reactivity feedback mechanis ms, recognized studies and experimental heat transfer DNB
(b) at least six delayed neutron groups, (c) both axial data. A minimum DNBR should be determined from the and radial segmentation of the fuel element, (d) coolant evaluation of the experimental data to ensure a 95%
flow provision, and (e) control rod scram initiation on probability with a 95% confidence level that DNB has either coolant system pressure or neutron flux. not occurred for the fuel element being evaluated. One example of a correlation which has been used to date is
12. The analytical models and computer codes used given by Tong (Ref. 7). The use of this correlation'and the should be documented and justified and the con above probabilities and confidence level yields a mini servatism of the models and codes should be evaluated mum DNBR of 1.30. Other DNB or clad failure both by comparison with experiment, as available, and correlations may be used if they are adequately justified with more sophisticated spatial kinetics codes. In par by analytical methods and supported by sufficient ticular, the importance of two- or three-dimensional flux experimental data.
REFERENCES
I. A. F. Henry, "Computation of Parameters Appear 3. R. A. Hein and P. N. Flagella, "Enthalpy ing in the Reactor Kinetics Equations," WAPD-142, Measurements of U0 2 and Tungsten to 32600 K,"
December 1955. GEMP-578, February 1968.
2. G. E. Moore and K. K. Kelley, "High Temperature 4. L Leibowitz, L. W. Mishler, and M. G. Chasanov, Heat Contents of Uranium, Uranium Dioxide and "Enthalpy of Solid Uranium Dioxide from 25000K
Uranium Trioxide," Journal of the American to its Melting Point," Journal of Nuclear Materials, Chemical Society, Vol. 69, p. 2105, 1949. Vol. 29, pp. 356-358, 1969.
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5. M. G. Chasanov, Argonne National Laboratory, Oxide," Nuclear Science and Engineering, Vol. 8, Reactor Development Progress Report, ANL-7618, pp. 497-506,1960.
September 1969.
7. L. S. Tong, "Prediction of Departure from Nucleate Boiling for an Axially Non-Uniform Heat Flux
6. E. Hellstrand et al., "The Temperature Coefficient Distribution," Journal of Nuclear Energy, Vol. 21, of the Resonance Integral for Uranium Metal and pp. 241-248, 1967.
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APPENDIX B
RADIOLOGICAL ASSUMPTIONS
The assumptions given below should be applied in pressure defined in the technical specifications for determining a conservative source term and subsequent containment leak testing.
transport of activity and resulting doses to the public for use in evaluating the radiological consequences of a I. Release of fission products to the secondary control rod ejection accident. system should be computed by assuming that all fission products released from the fuel clad are uniformly
1. The assumptions related to the release of radioactive mixed in the primary coolant volume.
material to the primary containment are as follows:
j. The primary-to-secondary leak rate limitation a. The case resulting in the largest source term incorporated or to be incorporated as a technical should be selected for evaluation. specification requirement should be assumed to exist until the primary system pressure falls below the b. The nuclide inventory in the fuel elements secondary system pressure.
potentially breached should be calculated, and it should be assumed that all gaseous constituents in the fuel-clad k. The release of fission products from the gaps are released. secondary system should be evaluated with the assump.
tion of a coincident loss of offsite power.
c. The amount of activity accumulated in the fuel-clad gap should be assumed to be 10% of the iodines and 10% of the noble gases accumulated at the end of 2. Acceptable assumptions for atmospheric diffusion core life, assuming continuous maximum full power and dose conversion are:
operation. a. The 0-to-8-hour ground-level release concentra d. No allowance should be given for activity decay tions may be reduced by a factor ranging from one to a prior to accident initiation, regardless of the reactor maximum of three (see Figure 1) for additional dis status for the selected case. persion produced by the turbulent wake of the reactor building in calculating potential exposures. The volu e. The nuclide inventory of the fraction of the metric building wake correction, as defined in Section fuel which reaches or exceeds the initiation temperature 3-3.5.2 of Meteorology and Atomic Energy 1968 (Ref. 1),
of fuel melting (typically 2842fC) at any time during should be used only in the 0-to-8-hour period; it is used the course of the accident should be calculated, and with a shape factor of 1/2 and the minimum cross
100% of the noble gases and 25% of the iodine sectional area of the reactor buiding only.
contained in this fraction should be assumed to be available for release from the containment. b. No correction should be made for depletion of the effluent plume of radioactive iodine due to deposi f. The effects of radiological decay during holdup tion on the ground or for the radiological decay of in the containment or other buildings should be taken iodine in transit.
into account.
c. For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of a
& The reduction in the amount of radioactive person offsite should be assumed to be 3.47 x 10-4 material available for leakage to the environment by m3 /sec. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the containment sprays, recirculating filter systems, or other breathing rate should be assumed to be 1.75 x 10-4 engineered safety features may be taken into account, m3 /sec. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the end of the accident, the but the amount of reduction in concentration of rate should be assumed to be 2.32 x 10- m3 /sec.
radioactive materials should be evaluated on a case-by (These values were developed from the average daily.
case basis. breathing rate [22 x 107 cm3 /day] assumed in a report (Ref. 2) of ICRP. )
h. The primary reactor containment should be assumed to leak at the leak rate incorporated or to be d. The iodine dose conversion factors are also given incorporated as a technical specification requirement at in Reference 2.
peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50%
of this leak rate for the remaining duration of the e. External whole body doses should be calculated accident.' Peak accident pressure is the maximum using "infinite cloud" assumptions, i.e., the dimensions of the cloud are assumed to be large compared to the t
distance that the gamma rays and beta particles travel.
The effect on containment leakage under accident "Such a cloud would be considered an infinite cloud for conditions of features provided to reduce the leakage of radioactive materials from the containment should be evaluated 2 on a case-by-case basis. International Commission on Radiological Protection.
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a receptor at the center because any additional [gamma g. The atmospheric diffusion model should be as and] beta emitting material beyond the cloud dimen follows:
sions would not alter the flux of [gamma rays and] beta particles to the receptor." (Ref. 3) Editorial additions were made to the quotation so that gamma as well as beta (1) The basic equabion for atmospheric dif emitting material could be considered. Under these fusion from a ground-level point source is:
conditions, the rate of energy absorption per unit volume is equal to the rate of energy released per unit XIQ *OyOz volume. For an infinite, uniform cloud containing X where curies of beta radioactivity per cubic meter, the beta dose in air at the cloud center is:
X = the short-term average centerline value of
= 0.457"E9
=D' the ground-level concentration (Ci/m 3 )
Q = amount of material released (Ci/sec)
The surface body dose rate from beta emitters in the u windspeed (m/sec)
infinite cloud can be approximated as being one-half this oy = the horizontal standard deviation of the amount (i.e., PD` = 0.23 Tp,. For gamma emitting plume (meters), [see Figure V-1, Ref. 5].
material, the dose rate in air at the cloud center is: z = the vertical standard deviation of the plume (meters) [see Figure V-2, Ref. 51.
S=
(2) For time periods greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the From a semi-infinite cloud, the gamma dose rate in air plume should be assumed to meander and spread is: uniformly over a 22.50 sector. The resultant equation is:
7D' = 0.25 EY 7x
- dQ =2.032 GzUX
where where PD' = beta dose rate from an infinite cloud x = distance from the point of release to the (rad/sec) receptor; other variables are as given in
,yD; = gamma dose rate from an infinite cloud paragraph g. (1), above.
(rad/sec)
= average beta energy per disintegration (Mev/dis) (3) The atmospheric diffusion model3 for
= average gamma energy per disintegration ground-level releases is based on the information in the (Mev/dis) table below.
X = concentration of beta or gamma emitting isotope in the cloud (Ci/m 3 )
Time f. The following specific assumptions are accep Following table with respect to the radioactive cloud dose calcula Accident Atmospheric Conditions tions:
0-8 Pasquill Type F, wind speed I m/sec, uniform
(1) The dose at any distance from the reactor hours direction should be calculated based on the maximum concentra tion in the plume at that distance, taking into account 8-24 Pasquill Type F, wind speed I m/sec, special meteorological, topographical, and other char hours variable direction within a 22.50 sector acteristics which may affect the maximum plume con centration. These site-related characteristics must be 1-4 (a) 40% Pasquill Type D, wind speed 3 evaluated on a case-by-case basis. In the case of beta days m/sec radiation, the receptor is assumed to be exposed to an (b) 60% Pasquill Type F, wind speed 2 infinite cloud at the maximum ground-level concentra m/sec tion at that distance from the reactor. In the case of (c) wind direction - variable within a 22.50
gamma radiation, the receptor is assumed to be exposed sector.
to only one-half the cloud owing to the presence of the ground. The maximum cloud concentration always
3 should be assumed to be at ground level. This model should be used until adequate site meteorological data are obtained. In some cases, available
(2) The appropriate average beta and gamma Information, such as meteorology, topography, and geographical energies emitted per. disintegration, as given in the Table location, may dictate the use of a more restrictive model to of Isotopes (Ref. 4), should be used. insure a conservative estimate of potential offsite exposures.
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Time. Time Following Following Accident Atmospheric Conditions Accident Atomospheric Conditions
4-30 (c) 33.3% Pasquill Type F, wind speed 2 days m/sec (d) Wind direction - 33.3% frequency in a
4-30 (a) 33.3% Pasquill Type C, wind speed 3 22.50 sector.
days m/sec (4) Figures 2(A) and 2(B) give the ground-level (b) 33.3% Pasquill Type D, wind speed 3 release atmospheric diffusion factors based on the m/sec parameters given in paragraph g&(3), above.
REFERENCES
1. D. H. Slade, ed., "Meteorology and Atomic Energy 4. C. M. Lederer, J. M. Hollander, and I. Perlman,
- 1968," TID-24190, Division of Technical "Table of Isotopes," Sixth Edition, University of Information, U.S. Atomic Energy Commission, July California, Berkeley, Lawrence Radiation Lab
1968. oratory.
2. Report of Committee I1, "Permissible Dose for Internal Radiation," ICRP Publication 2, 1959. 5. F. A. Gifford, Jr., "Use of Routine Meteorological Observations for Estimating Atmospheric Dis
3. D. H. Slade, op. cir., Section 7.4.1.1. .persion," Nuclear Safety, Vol. 2, No. 4, June 1961.
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