ML20064M084

Forwards Amended PSAR Chapter 4 & 5 Re Gas Entrainment in Primary Heat Transport Sys.Info Will Be Included in Next PSAR Amend

ML20064M084
Person / Time
Site:	Clinch River
Issue date:	02/11/1983
From:	Longenecker J ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:	Grace J Office of Nuclear Reactor Regulation
References
HQ:S:83:209, NUDOCS 8302150457
Download: ML20064M084 (11)

Text

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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:83:209 Dr. J. Nelson Grace Director FEB 11 m3 CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ,

Washington, D.C. 20555

Dear Dr. Grace:

ADDITIONAL INFORMATION CONCERNING GAS ENTRAINMENT IN THE PRIMARY HEAT TRANSPORT SYSTEM (PHTS)

Enclosed are amended Preliminary Safety Analysis Report (PSAR) pages to Clinch River Breeder Reactor Plant PSAR Chapters 4 and 5 that provide additional information concerning the potential for gas entrainment in the PHTS. These pages are in response to questions raised by the Nuclear Regulatory Commission staff reviewer and will be included in the next amendment to the PSAR.

Questions regarding the enclosure may be addressed to Mr. D. Hornstra (FTS 626-6110) or Mr. D. Robinson (FTS 626-6098) of the Project Office Oak Ridge staff.

Sincerely, l &*

John R. Longenecker Acting Director. Office of Breeder Demonstration Projects Office of Nuclear Energy Encicsure cc: Service List Standard Distribution Licensing Distribution 9#

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ENCLOSURE the Doppler feedback in CRBRP. Reducing th'e Reference 9 uncertainty for the already conservative neglect of global temperature-importance .

weighting results in an extrapolated Doppler uncertainty of less than ,

110% (la) for CRBRP.  !

' Further confirmation of the accuracy cf.LMFBR Doppler constant '

--- predictions is provided by the small-sample measurements in zero power l critical mockups. Small, heated-sample Doppler constant measurements '

have been performed in the. Zero power plutonium Reactor (ZPPR) critical assemblies simulating the CRBRP core configuration. The analysis of the .

Doppler constant experiments in IPPR-2, 3, and 5 are reported in i

l References 10, ll, and 12, respectively. Although these small-sample '

measurements do not represent a direct experimental detennination of the '

• total core Doppler constant .the good agreement between calculated and measured values .does provide substantial confidence that the U-238 ,

resonance parameters in the core spectrum are accurately predicted.

l From the total of 52 small-sample Doppler measurements throughout the core under a variety of reactor conditions (flooded, voided.. control '

rods inserted and withdrawn, etc.), the mean calculation-to-experiment ratio is 0.98 (slightly conservative underprediction of Doppler) with

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an uncertainty of 15.4% (la). . .' ,

• The CRBRP Doppler constant uncertainty, based principally on  !

the aforementioned SEFOR evaluation, and supported by the small-sample measurements in ZPPR, is + 10% (la). This value is intended for use in

{ operationel'and duty cycle.

design transient evaluations within the reacto

- and the extf emely unlikely class of events are based on Doppler constants with -2o (80% of nominal Doppler) and -3o (70% of nominal Doppler) uncertainties, respectively.

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4.3.2.3.2 Sodium Void Worth 4 g4g The sodium void worth relates the change in neutron Small, multipli-distribu-cation to the presence of voids in the sodium coolant.

ted voids such as gases entrained in the coolant 3are adequately treated by use of the sodiur, density coefficient (Section 4.3.2.

• l unlikely situation. In the following discussion, the reactivity associated with this latter type of voiding is developed for use in accident analysis in Chapter 15.

Figure 4.3-28 is a flow chartCross showing the are sections method for calculating processed both the sodium voiding reactivity worth.

~with sodium and without sodium to properly account for resonance self-shielding and the change in spectrum 1 when m :.sodium

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The above discussion refers to demonstrating the ability of the rods to meet their design requirements when subjected to the various l

- categories of design transients. Such topics as transient effects due to rod itilures and continued operations '

with fa'iled rods- are presented

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iri S~ection 15.4.

• 4.4.3.7 Potentially Damaging Temperature Effects During Transients i

Ivi general, a single anticipated event is not damaging to the reactor structu'res; it is the total sum of all the occurrences (in this -

event classification) over the particular lifetime that may cause the -

- structure to approach its design limit.- As mentioned in Section 4.4.2.9, the current cladding requirement for fuel pins is that considering all ..*

- normal'and anticipated events, the cumulative cladding damage must not l preclude the capability to survive at least one of the wprst unlikely -

events without loss of cladding integrity. .

Maximum 3c fuel and blanket hot rod temperatures during v'rious a limiting core design events of the plant duty cycle which are described i in Appendix B have been analyzed. Detailed temperatures for various '

axial and radial positions along the rods have been evaluated in deter-mining the core design adequacy as described in Section 4.2.1. Con-servative assumptions used in calculating these temperatures are described in Section 15.1.4. These assumptions include: full power operation, hot rod analyzed in highest power and temperature fuel assembly of all core

- conditions, worst case Doppler coefficient with uncertainties included, a 200 millisecond delay between trip signal and the start of control rod insertion, 3a hot channel factors, the single most reactive control rod assumed to be stuck in the withdrawn position for both sets of control i

rods, highest. core pressure drop, and.the most rapid flow coastdown of the primary pumps following pump trip. In addition. Section 15.1.4 shows

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that of the safety related events, both of the overpower and undercooling l

58} type, the Safe Shutdown Earthquake (SSE) event (60c step reactivity insertion occurring under SSE conditions) results in the highest core temperatures.

The conclusion of the Chapter 15.0 safety evaluations and the Section 4.2.1 design evaluations is that the design changes incurred in going from the homo-geneous to the heterogeneous scheme are not expected to significantly change the design or safety capability of the core.

D@ N.4.3.8 Thermal Description of the Direct Heat Removal Service (DHRS)

The thermal description of the Direct Heat Removal Service will be found in Section 5.6.2. .

55l 4.4.4 Testing and Verification At the present state-of-the-art in reactor design, scale model 51 tests of reactor flow systems provide the most useful tool for studying reactor hydraulics. pressure drop through complicated flow paths, thermal (i C 4.4-75 Amend. 58 Nov. 1980

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An evaluation has also been made to determina if expected levels of gas entrained in the sodium could impair heat transfer in the reactor.

It was found that at the limiting value of one volume

- - - percent established for reactor neutronics effects (section The

- 4.3.2.3.2) there is no significant effect on heat transfer.

expected entrainment. level (Section 4.4.4.1) is much lower than this limiting value.

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biet plenum as a function of the plenum height, Inlet nozzle discharge angle, and the elevation of the nozzles relative to the Core Support Structure. The test served a scoping purpose for use in designing the 1/4-scale IPFM test previously described.

The gross flow patterns were predictable and produced no unusual or unexpected -

results. - ,

.~' Intaa/al'Reac.ter Flow Model. Phase T Testtr.u - ou+Ie+ Planum Featura -

,fjow and Vibration Test -

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This test: 1) measured the velocity pattern in the outlet plenum and in the vicinity of the major outlet plenum structures; 2) determined the pressure -

drop characteristics cf the outlet plenum and major outlet plenum structures;

3) determined the mixing characteristics and transport times in the outlet plenum and at locations of probable hot / cold interf aces; 4) evaluated flow induced vibration characteristics of selected outlet plenum structures; and 5) . .

• evaluated the gas entrainment characteristics of the suppressor plate.

&SMWIntearal Reactor Flow Model. Phase Il Testina *

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The purpose of this te~st which is ln progress, is to verify the final design for hy.draulic and vibration performance using the Integral reactor flow model. ,

j Those components used in the Phase I testing, whose design has changed to the extent that hydraulic, and vibration performance is influenced, will be modified in a subsequent Phase 11.

Outlet Plenum Flow Stratification Test ,

. The Outlet Plenum Flow Stratification Test was performed in a model of 0.55 scale simulating a 120 sector of the CRBRP reactor vessel outlet plenum, containing a portion of the UIS, and an outlet main coolant pipe.

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Sane important conclusions and observations derived from the evaluation of the test data (Ref. 60) are: ,

1) The transient temperature response at the outlet nozzle is less severe than that predicted by the plant simulation model in the DEM5 code, thus demonstrating that this UlS design goal is satisfied.

2) The standard height Uls chimneys resulted in less severe transient -

temperature ranp rates at the outlet nozzle than shorter chimneys.

3) The nominal prototypic gep beneath the UlS skirt of 1.0 inch resulted in less severe outlet nozzle transient temperatures than either larger cr smaller gaps.

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A.4-77 Amend. 58

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INSERT FOR FAGE 4.4-77 The gas entrainment tests were performed to determine the amount of cover gas that might be entrained at the free surface inside Using conservative Weber Number the CRDRP reactor vessel.

podeling of the entrainment process, the average reactor vessel outlet void fraction when at normal liquid level and scaled to I 115 percent of CRBRP average full flow is 1.25 a 10-4 The paximum void fraction measured at any reactor vessel outlet nozzle for this same case and extrapolated to 115 percent flow l was 3.0 x 10-4 In addition to determining primary entrainment pheno.sena within the reactor vessel, tests were run to determine the o9uilibrium PHTS void fraction that results for a given gas entratament rate in the reactor vessel. These tests were based on the conservative assumptionThis thatassumption all gas removal occurs results in in highest the the out1ct plenum of the reactor.

equilibrium void fraction in the PHTS, Based on the entrainment rate that gave the void fraction of 1.25 x 10-4 and Weber Number modeling of the entrainment proceta, the equilibrium void fraction Gasin the PHTs removal at 115 by the percent IHX vent and oftherated primaty flow is 2.0 x 10-3 pump standpipe bubbler will reduce this invel.

The test results, as well as a complete description of the subject testing, are presented in Reference 76.

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73. F. C. Engel'. R. A. Markley and B. 'Minushkin. " Buoyancy Effects on Sodium Coolant Temperature Profiles Measured in an Electrically . . .

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. Heated Mock-up of a 61-rod Breeder Reactor Blanket Assembly.",, ~'

'ASME 78-WA/HT-25.' .

R. A. Markley and B. Minushkin. " Heat Transfer T'est 74.~ F.. C. Engel .

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' Data of a 61-rod Electrically' Heated LMFBR Blanket Assembly Mor.kup' '

and their use for.Subchannel Code Calibration" in Fluid Flow and * ~ ~

.- . Heat' Transfer Over' Rod or Tube Bundles symposium of ASME 1979 Winter Annua.1 Meeting. pp. 223-229 December 1979. .

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75. F. C. Engel. 8 Minubkin R'.' J.' Atkins and R. A. Markley. "Characteriz-

• ation of Heat Transfer and Temperature Distributions in kn Electrically -

Heated Model of an LMFBR Assembly, to be published in a special issue 58 . of Nuclear Engineering Design. -

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Gas entrainment within (11 ft/sec) at design flow frem the reactor vessel.

' the pump is minfalzed by designing and testing, to ensure that alI pump parts >

which need to be submerged during operation at pony motor speed or restart to pony motor speed are located below the alntmum sodium level.

jsed -PDe max! mum oxygen content in the primaryThis

• system level ofsodlum sodiumisimpurity specifled to be 42 -

ppm at 800*F or above and 35 ppm below 800 F. -- ,

will not ,. affect the pump operating characteristics.

The biolog! cal shielding for the PHTS sodium pumps as well as the pump .

assembiles is designed to withstand the loadings associated with the Safe . -

. Shutdown Earthquake (SSE) and the transient overpressures. for ex unlikely plant conditions. -

the PHTS pump as a Class I component in accordance with the rules of the ASME )

Code Section 111 and modifying RDT Standards. . '

I The biological shlalding for the PHTS sodium pumps is p'rovided by (1) an f annular shield tank which surrounds the pump shaft, (2) the pump shaf t itself l' which is designed to provide an Integral part of the shield requirements, (3) the pump support structure which is part of the operating floor, and (4) l special precautions to preclude streaming along the Instrumentati penetrations. flange of the pump pressure boundary containment vessel which accordance with the ASME Code for Class I nuclear components.

supporting assembly and the annular shield structura The design of this joint provides for the operating floor pump motor w' ell. i 1 ds and provides a seal at

' .the dual function of resisting static and dynam c .oa the boundary.between the pump atmosphere and the ,

rings.'

The SSE For the PHTS pump SSE. seismic analysis, a 2$ damping value Following was used.

loadings were considered to occur in conjunction with a plant trip.

the SSE, the Intermediate Heat Transport System, Steam Generator Syste Steam Generator Auxillary Heat RemovalComputer stored and decay heat.

System programs, must provi flow without loss of structural Integrity after the SSE.such Descriptions of these computer programs can be found in primary pumps.

Appendix A. .

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5.3-43 . . . . . <<.

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INSERT FOR PAGE 5.3-43 The pump standpipe bubbler also provides a vent for gas de-entrained fro:n the PETS. The maximum gas flow that the '

. . standpipe can accommodate without flow instability corresponds to an entrained gas volume fractionef 3 x 10-4 This volume fraction has been shown to have no impact on pump performance or heat transfer in the IBX. Outlet plenum tests which give expected gas entrainment levels are discussed in Section 4.4.4.1.

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i 5.3.3.4 Valve- Cha racterisitics The essential characteristics and features for the check valve i

design to satisfy the system operating requirements are included in Table 5.3-13.

The CRBRP Cold Leg Check Valve (CLCV) is hydraulically similar to the FFTF 16" check valve. As noted in Reference 62, extensive tests have been performed on the FFTF valve to confirm its hydraulic performance. In operation at FFTF, no problems due to gas accumulation in the CLCV have been identified. Secondly, tests performed on the FFTF hot leg isolation valve verified the self-venting characteristics (that any accumulated gas will be swept out) of the main valve body of both valves. Since the valve bodiee of both CRBRP and FFTF CLCV's and the FFTF hot leg isolation valve are geometrically equivalent, the results of these tests are applicable to the CRBRP CLCV. Air deliberately introduced into the valve was quickly entrained and removed by the turbulent fluid at pipe flow velocities greater than 7 fps, which correspond to a 2 fps (average) velocity in the valve. The experimenters noted that, "This is the approximate limiting value for air removal from pipelines as established by previous research" (Reference 63). Under full flow conditions the flow velocity in the CLCV inlet pipe is about 25 fps and about 6 fps (average) in the valve body. At 40 percent flow the velocity in the valve body is still above the 2 fps level, and at this reduced flow rate there is little likelihood for any gas entrainment at the principal gae entrainment site (the reactor vessel outlet plenum). Thus, no gas accumulation is expected in the main body of the CLCV.

Although no mechanism has been identified which would allow the de-entrainment of gas in the valve, an evaluation has been performed with the following conservative assumptions. The dome on the top of the check valve is initially ful- gas at the conditions consistent with the thermal hydraulic os_ign condition, the pumps are tripped, and the pressure in the valve has been minimized by assuming that the sodium level in the reactor is at the minimum safe level. The evaluation showed no break of siphon. Therefore, gas accumulation in the check valve would not be a problem for CRBRP.

5.3-43b

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References:

54. P. Soo, "Seiection of Coolant-Boundary Materials for the Clinch River -

Breeder Reactor Plant", WARD-0-0010, August 1974. .

55. Deleted

56. Deleted

57. P. Soo (Compiler), " Analysis of Structural Materials for LWBR Coolant-Boundary Components - Materials Property Evaluations", WARD-3045T3-5, November 1972.

58. Deleted .

59. R. A. Leasure, "Ef feet of Carbon and Nitrogen and Sodium Environment on the Mechanical Properties of Austenitic Stainless Steels",

WARD-NA-94000-5, De'2 ember 1980.

60. ES-LPD-82-007, CRBRP Engineering Study Report, "CERP Transition Joints", April 1982.

61. ES-LPD-82-008, CERP Engineering Study Report, "CERP Materials Data Base", May 1982.

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