ML19344A336

Monthly Operating Rept for Mar 1971

ML19344A336
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Site:	Midland
Issue date:	04/15/1971
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Download: ML19344A336 (16)

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- TABLE OF CONTENTS Page

-A. LOFT Integral Experiments and Radiological Studies . . . . . . . . . 1 B. Loss-of-Coolant Accident Analysis'. . . . ... . . . . . . . . . . . . 3 C. Technical Assistance in Reactor Safety Analysis . . . . . . . . . . . 12

-D..* Separate Effects Testing . . ........_............ 14 r

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March.1971' s .

HIGHLIGHTS

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-Ten. LOFT Project design documents _ and ; twenty-five sections of. the LOFT Project Design Basis-Report were reviewed.

' The' document tree setting forth LDFT program and test planning documents was approved with commentL by AEC. A' schedule' for preparation of these _ documents i .

was submitted to AEC.

1 RELAP3 predictions of Semiscale Blowdown Tests 848 and 850 with and '

without emergency-core-coolant'-(ECC) injection, respectively, were made.

Agreement:between predicted and measured pressures was good; the maximum difference was 100 psi.- The predictions and. experimental data both: indicate that in the Semiscale system, ECC injection does not have a marked effect on system' pressure decline-.

The recently. completed series of'Semiscale tests have indicated some deficiencies in our analysis. . Efforts are continuing to idectify-and resolve these deficiencies. Semiscale is'sufficiently complete and meaningful as a one dimensional ~ representation of the . blowdown and ECC interaction processes that it can provide' a' measure of the adequacy of the analysis techniques to -

characterize _these, processes in one dimension for safety system assessment purposes. -

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The document describing THETAl-B single rod, single' channel code for core ' thermal analysis was : published.

Eight papers were presented _ at the National Topical Meeting of the American Nuclear So'clety-at Idaho Falls, Idaho, March 29-31.

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, March 1971

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'A. LOFT Integral Test Program Development

1. Highlights Ten' LOFT Project design documents and twenty-five sections - of the LOFT Project Design Basis Report'were reviewed.

The document tree' setting forth LOFT program and test planning documents was. approved with comment by AEC. A schedule for preparation of these documents was submitted to AEC.

"2 . Technical Activities-The following LOFT Project design documents were reviewed for compliance with the requirements sett forth in PRD-1B, LOFT Program Requirements Document,

-and comments. were submitted to 'the LOFT Project.

(1)- CDD' 1.1.1.5A, " Core Support Structures" (2) SDD 1.1.3B, " Emergency Core Cooling System" (3) SDD 1.1.4D, " Primary Coolant System and Its Subsystems" (4)' SDD 1.21B, " Pressure Reduction Spray System" (5) SDD 1.2.3B, " Containment Vessel Railroad Door Operating and Sealing System"

-(6) SDD 1.2.8A,_ "Containmeat ' Isolation System" (7) SDD 1.2.].4A, '"Persennel Airlocks System"

• i (8) SDD 1.2.17A, " Piping Penetrations Through containment Vessel System" (9)- SDD 1.3.3A, " Demineralized Water and Softened Water System" (10) SDD -1. 3.26A, " Railroad System" -

.The following ~ sections :of the LOFT Project Design Basis Report were-

. reviewed and commented on:

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-(l)-~Section-1-A,." Purpose of Report" L

-(2) Section I-B, ' " Mission ' of LOFT Program" (3) . - Section I-C. . " Design Objectives" s

-(4) - Section I-D,7 " Summary of the Method used to . Develop LOFT Reactor Plant Design" -

(5) Section1II-A, " Description-of LOFT Plant"'

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'- -- (6)= Section:II-B, " Identify Additionci Support Work" ,

• . - (7). Section :III-~A',L " Description of.. LOFT Background"

-(b) Section.III-B, " Facility-Description" l

!. (9) ' Section III-C,; " Reactor Plant General Design . Features" (10) - Section IV-A-l',l" Reactor Power Leva.1"

- (11) Section 'IV-A-5, _ " Core Flow Resistance" (12)! Section IV-A-9, :" Initial Core Coolant Velocity" (13) Section IV-A-12, " Initial System T,y," '

(14)nSecti$n IV-A-13, ~ " System Pressure"

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(15) Section IV-A-14, Initial Core . AT" (16) Section;IV--A-15, " Coolant System Component Volumes and Distribution"

. (17) Section IV-A-17, " Core Flow Area / Break Aree Eatio"

~(18) Section IV-A-18, " Coolant System Component Elevations" f(19)JS'ection IV-A- 19, " Initial Coolant System Flow. Velocity" (20) Section IV-A-20, " Containment Back Pressure" '

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(21) Section. IV-A-21, "ECCS Injection Locations"

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[ (22).lSection IV-A-22, "ECCS Inj ection Saturation' Points" l

[. (23) Section .IV-A-23, "ECCS Injection Characteristics"

-(24)~Section'IV-A-26, "Effeet of Secondary Coolant System" '

'(25).Section IV-C, " Additional Work to Substantiate Design" l

!' Preparr ' *on ~of- the . LOFT Technical Activities Description for FY-72 commenced. _

- The dScument : tree . setting forth'- LOFT program and test planning documents

was approved with comment by - AEC. -A schedule for preparation of these documents E

-was submitted to;AEC. 'In this.' group,l Experimental Requirements Document and'

Measurements ? Requirements Docur nt were previously submitted 'to AEC. for comment.

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' Informal comments :have been rec ~eived from AEC and -have' been the subject of idiscussions with them.

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March 1971-B. ' Loss-of-Coolant Accident Analysis

1. Highlights-RELAP3 predictions of'Semiscale Blowdown Tests 848 and 850 with and without . emergency-core-coolant _ (ECC) injection, respectively, were made.

Agreement ~between predicted and measured pressures was good; the maximum difference was 100 psi. The' predictions and experimental data both indicate that;in the.Semiscale system, ECC injection does not -have a marked effect son system pressure decline, i

The recently. completed series of Semiscale tests have indicated some

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deficiencies'in our 'nalysis.

a Efforts are: continuing to identify and resolve

.these deficiencies. Semiscale is sufficiently complete and meaningful as a one dimensional representation of the blowdown and ECC interaction processes

-that it'can provide a measure of'the_ adequacy of the analysis techniques to l-characterize these processes in one dimension for safety system assessment purposes, i

l The document describing THETA 1-B single rod, single channel code for l core themal analysis was published (l).

l The following-papers were presented at the National Topical Meeting of the American Nuclear Society at Idaho Falls, Idaho, March 29-31.

Idaho Nuclear Code ' Automation: A Standardized and Modularized l Code Structure, C. W. 'Solbrig, L. J. Ybarrondo, R. J. Wagner.

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j. Migration,of Fission Products in Operating UO2Fuel Rods, W. A. Yuill, V. F. Baston, J. H. McFadden.

MOXY: A Digital Computer Code for Computing Core Haat Transfer,

• D. R. Evans.

Mathematical Techniques for Vibration Analysis of Reactor

- Systems, C. Noble. *

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Flow Simulation Through Graphics: A High-Speed Interpretation, M.-A. Lintner, C. W.'Solbrig.

Nodal-Sensitivity in Modeling a Large Power Water Reactor System for a Loss-of-Coolant' Accident, C. E. Slater'and R. D. Hentzen.

Effects of State. Properties o1 High Pressure Compressible Hydro-dynamics, K. V. Moore, G. E. Gruen.

(1) (C. J. Hocevar and T..W. Wineinger, THETA 1-B,' A Computer Code for Nuclear Reactor Core Thermal Analyses, IN-1445 (February 1971).

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,. ~(Hydraul'ic Modeli Development: RELAP3 predictions ' of ' system dec'ompression a ;wer'e obtained; for Semiscale -Blowdown Tests 848 and .850.~ The' test conditions for ~ these ' tests were; similar, .with j thNmajor= exception being :that ECC was cinjected .(into ; the : lower plenum) ;during Test 848 'but ' not_ during Test- 850.

The experimenta1rsystem used.for the' blowdown testsLis_shown in Figure _1.

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. . . The RELAP3L.model ofLtheLsem'isca'leisystem isishown in Figure 2. The-predictions Land experimental pressure '_ data 'at- the vessel outlet nozzle

'for' Test.:848Jareicompared'inFigure 3. lAs shown in Figure 3, before ECC injection, the' maximum deviation"between the' experimentai data and _ .

analytical predictions,is1about 100_ psi.' Shortly-after injection the devia--

tion is.-largerfandiconsideredjto beidue to the factfthat RELAP3 calculates-

-the mixing (of the:ECC withLthe fluid'in the:. system _as.a-homogeneous process.

-TheLexperimental. data'from Teses!848 and.850!shown in Figure 4. indicate mixing 1to be; poor..1If,the experimental error:(+ 50 psi) and the-differences in. initialj conditions Lareidiscounted, the prescure behaviors : are quite

similarE for thej testsk ;The .similarityJ indicates that Ithe ECC injected c into lthepsystem didLnot mix significantly with the system fluid and had

-little effection _ depress'urization behavior.

TheLeffect' of the homogeneous mixing assumption in RELAP3 on predicted depressurization behavior is evidenced in.the comparison lof' pre'dicted pressures for Tests 848 and-850, shown in Figure 5.-

The purpcse of the. current; data analysis and interpretation tasks is to' identify potential improvementsLinLthe: current' analytical models so that

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they willJb'e capable of predicting the behavior' observed in'the-semiscale tests and. LOFT.1 1 ,

ItLshould be noted-that the.recently completed' series of semiscale- '

' test'sTprovide data:bn(the primary ' system Jdecompression -history with the

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11nteraction of accumulator, ECC fluid. This system includes the najor features cd ' '

'of the; primary coolant; circuit and the initial fluid ecargy: distribution ,

characterizing a'LPWR:with the qualification that the multiloop aspects of a

- 1LPWR are.representedibj a single? loop-in'semiscale. -Further the simulated i 'pipejrupture?in'semiscale. characterizes laEsimultaneous pipe break in all:

Lloops l(iej 2,f3p orf4) of.'a LPWR withja break size equivalent to breaking -

.one inletJpipe'ofla;fouriloop-plant. JThus the semiscale system is a t

.x . single { loop representation 1of.alLPWR with one dimensionaliflow and as such -

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connected directly3tolthe4 ower plenum.- : Since"in large systems accumulator.

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einjection isiatia .commonllocation with the ; inlet, the ECC in semiscale'is finjected"directlyJined the lower plenum.

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' :which Ts representative offlarge system l behavior. One of the purpos~es of.the  !

Jsemiscaleftest's has been:to provideran experimental basis'for assessing the 1

[ ' 1adequacylof(ourf analyses.. cIn their! present istate the ~ analyses do not provide'

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~l e 'an,adequateidescriptionTof?the phenomenaLobserved in semiscale.. Thus they' cannot (at thisistageEbe usedlas ~ a' basis of confidence for other similar :

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~ processes,(thatilt can provide a: measure'of the adequacy of the analysis fechniqu'esito characterize;these' processes;in one' dimension-:for-safety

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. system assessment purposes.1 Core-Therma Mode 1' Development:.' Development-o'f the THETA 1-B code is

. continuing. ; ' An3 automatic procedure fo'r taking ~ RELAP3 output and preparing-

. it automatically;.as THETAl- B (inputi was modified to permit operation with' multiple inputitapes.1 The ECCSA-4 and MUCHAl codes of BMI, as-well'as the THEAT1-B' code,;were compared and differences; evaluated using the Idaho

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7 sNuclear Corporation' code classification systems which was developed for this^ purpose.'

Effort has' continued on developmenta.td modification of the core heatup 2

code,:M0_XY,'Tto permit calculation of 'the temperatures .of the core instrument'

'supporti tubes;during the planned " LOFT loms-of-coolant experiments. Thermal radiation' view' factors'willibe entered into MOXYLalong with the necessary-geometry and. logic changes before the calculation can be ' accomplished.

Structural Model. Development: A. modification was made to the STRAP-S computer code. .Thennew version of the code',l STRAP-S/ MOD 001 was. generated to correct.an-error' associated with released deg'rees of freedom and an error Lin the~ transformation of membersiinto the structural coordinate system.

AItechn h al paper?" STRAP --A Computer. Code for.the Static and Dynamic.

Analysis .of lReactorf Structures":was -transniitted = to Berlin, Germany, for.

publication :in ithe proceedings _of the First' International Conference on

.'Struetural: Mechanics 11n' Reactor Technology'.';

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'y A code; to solve the' complex;eigenvaltie problem l A-ABl in which B is

'nonsingblar was1 completed. . The present program .will handle matrices 'up

-to 10 7by110;(however, it;can be madeito handle' larger matrices if the -

situat' ion arises. = l The program transforms ? the matrices to ' Hessenburg form -

b'y.~ Householder transformations and the';resulting' matrices are solved by the -QR methdd with origin shif ts. -

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A' computer; program was written'toidetermine the effects of earthquakes 4: 'on:tanksJcontaining li uid.- The basic method of solution follows.that

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outlined in TID-7024(2 .- The. solution' indicates (1):the peak stress induced i

~ ' /in;theitankidurint the: earthquake ~,-(2) whether' the tank will buckle during (the " earthquake,: andz.(3)"whetheritheltank, if :notl anchored, will; tip over.

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.during;the' earthquake.

' The5 acceleration' histories of four<new earthquakes were added - to the -

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_ iearthquake; library forfstruc'tural dynamics. :The;following earthquake records- .

are.now?available inLa; form permittingLdirect input into the' STRAP-D computer s 1 program'

: ;(1)l1940; El Cen'tro,T (2)f19.43 El: Centro,- (3) 1952 Taf t, (4)'1949' W g olympia,7and L(5)il966 LParkfield.' J The"first' four earthquakes are considered

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.the.four strongest recorded in the USA. Seismic experts have suggested that the. earthquake resistance 'of nuclear reactor systems be measured by how the. systems respond to the ground motions of these earthquakes L(TID-25438)(3). This approach is the deterministic approach to seismic analysis.'

A computer program to generate artificial earthquake forcing functions is being developed. . The'. program'is now in the checkout stage. The program uses:the method suggested _by Housner and Jenninge 4) The earthquake .

ensembles generated by che' program will permit a probabilistic approach '

to the seismic analysis of nuclear reactor systems. Other seismic experts have contended that _this . approach is superior to the determinsitic approach 3) .

A method to determine'the frequency spectrum of earthquakes has been developed. .The method is based on the principles of Fourier analysis. The primary application of the method is to determine the dominant frequencies of earthquake ground l accelerations. The dominant earthquake frequencies

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can then be compared with the natural frequencies of structures. The-extent to which any particular structure will be excited by any particular earthquake can.thus be estimated quickly.

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Fission Product Behavior Anslysis: Observed surface temperatures'during irradiation have been correlated with preirrediation pin parameters, particularly the width of the fuel-cladding gap. Some difficulty has been experienced in obtaining satisfactory agreement among the operating surface temepratures reported by different experimenters for apparently identical pins. The'effect

. of increasing the fuel'-cladding gap from 6 to 12 mils fo'r BWRs is also being studied because an incregse in coefficients for_the gap t5,6). gap width willindecrease A decrease thetransfer the gap heat heat transfer coefficient will increase fuel surface temperature and the release of noble gas fission product's from the fuel.

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, (3) S. D. Werner, A Stu'dy of Earthquake Motions for Seismic Design, '

TID-25438 (June 1970) . .

(4) G. W. Housner, P. C. _ Jennings, " Generation of Artificial Earthquakes,"

-Proceedings of the American Society of Civil Engineers, Journal of 1the Engineering' Mechanics Division (February 1964)pp. 113-150.

(5)[ C.- N. ' Craig, G. R. _ Hull, W. E. Baily, -Heat Transfer Coefficients Between Fuel- and Cladding in Oxide Fuel Rods , GEAP-5748 (January 1969) .

(6) W. 'E. Baily,~ C. N. Craig, E. : L.' Zebroski, "Ef fect of Diametral Gap

.  ; Size on 'the In-Pile Performance of Fast Ceramic Reactor. Mixed-Oxide Fuel," Transactions 'of the American Nuclear Society, 9 - (l),

(June 1966) .-

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C.; ^ Technical Assiatance in Reactor Safety Analysis l i

1. Highlights-The paper " Correlations ' to Predict Maximum PWR. Containment Pressures Following: a foss-of-Coolant' Accident" by D. C. Slaughterbeck and H. Specter

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(AEC) was presented at the ANS Topical Meeting on New Developments in Reactor -

Mathematics and Applications, March ~30, 1971.

.2. Technical Activities-

' Containment Response Analysis: The vent clearing model in CONTEMPT-PS( } <

has'been reevaluated by comparing the calculated vent clearing times and the dry well pressures at the time of vent- clearing with those measured experimentally in Humboldt Bay Tests 15, 16, 17,and 23. These tests cover a range of break area-to-vent area ratios from 0.015 to 0.060. This range of ratios corres-ponde to breaks in the simulated reactor system ranging from 100 to 400%

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of the maximum credible operating accident breaks postulated for the Humboldt Bay design. . .

1 The mass and energy input rates from the reactor simulator $nto the dry well for Humboldt Bay tests were calculated with the RELAP(8/ code, because these rates were not measured experimentally. One set of RELAP3 runs was . made in which the input parameters were varied until the pressure histories and total blowdown times calculated by RELAP3 taatched those measured ~in the reactor simulator for Tests- 15, 16, 17,and 33. Although

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this procedure :gave the best overall fit to the measured pressure data

. during the complete course of the bJewdowns, it did not necessarily give the best fit for the early part of _ the transient (during vent clearing).

Since the rate at which mass and energy is added to the dry well has a strong -influence on vent clearing time, additional RELAP3 runs were made to more closely match or bound the experimental pressure during the vent

', clearing portion of the transient. These runs were.' performed by calculating <

the blowdown response through use'of the Moody critical flow model(9) and a discharge coefficient of one. The measured initial reactor simulator fluid saturation pressures or 1265,1000, 900 and 800~ psia were used for ~

each of the four-tests.

Calculated and experimental dry well pressures. at the. time of vent clearing are plotted' as a. function of the vent clearing time 'in Figure 6.

Shorter vent clearing times and lower dry well pressures were predicted than were observed for. the tests. The CONTEMPT-PS vent clearing model will be examined to determine, whether it can be improved. -

-(7) ' C. F. Carmichael and S. A. Marko, CONTEMPT-PS - A Digital Computer

. Code for Predicting the Pressure-Temperature History Within a Pressure-

' Suppression Containment Vessel in Response to a Loss-of-Coolant Accident, IDO-17252 (April 1969).

(8) -W. H. Rettig, et al. ,' RELAP3 - A Computer Program for Reactor Blowdown

. Analysis,- IN-1321 (June 1970).

(9) : F. J. Moody, " Maximum Flow Rate of a Single Component, Two-Phase

. Mixture," . Journal Heat Transfer - Trans. ASMF, '87 ml (February 1965).

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. e M rch-1971

.D. ' Separate Effects Testing

- 1. Highlights L Preliminary comparison of staiscale data for tests with and without emergency coolant indicate ECC injection causes only local disturbances and does not. significantly. affect overall system decompression behavior,

2. Technical Activities

' Single-Loop Semiscale: Analysis of the results of Seniscale Test 850, performed late in February (101 has been completed. The purpose of Test 850 was to obtain' base-line data without .ECC injection for comparison with data from a test whichL included emergency cooling. Test 850 was conducted with a high inlet break configuration from initial conditions ye similar to those for the ECC-injection Test 848, reported previouslyll . A preliminary comparison of the results - for 'the two tests indicates that ECC injection causes density, flow, and temperature disturbances near the injection point, but that these localized disturbances have no significant effect on overall system decompression behavior, as also discussed in Section B and illustrated by Figurc 4.

Semiscale Test 851 was performed with ECC injection into the lower plenum.

to determine the effect of emergency cooling'on the decompression behavior for an outlet break. Conditions for' Test 851 were the same as for Test 825(12) without ECC injection. ' Analysis of the data for Test 651 is in progress.

Dismantling of the in. - single-loop . system was completed in preparation for construction -of a 1-1/2-loop system that more closely resembles the LOFT system. The 1-1/2-loop system will include a ~ 66-ir.. heated core, the capability for accumulator ECC injection, one complete . operating primary loop, and separate piping for blowdown initiation.

. BWR.-F1ECHT: -Recent emissivity measurements made by the' Heat Transfer Division,.TRW, Inc.,-are significantly higher (up to 50%) than the values for the emissivity of stainless steel used by General Electric Company in analysis of the data from which their Rogers Model for top-spray cooling (73) was. developed. As a result, parameters used in the GE model for top spray .

cooling used for accident evaluation require modification.

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PWR-FLECHT: A ' draft of. the final rcport for the PWR-FLECHT Test Program was received from Westinghouse Electric Corporation. Idaho Nuclear Corporation ,

has. reviewed the report and submitted comments to Westinghouse.

. .(10) . Nuclear Safety Program . Division Monthly Report .- February (Hai-54-71),

narch ' ll, 19 71. ,

(11)' Nuclear Safety Program Division Monthly Report - January (Hai-2641 ),

. Feb ruary '12, 19 71.

(12) Nuclear' Safety Program Division' Monthly Report - June (Hai-269-70),

July 15, 197J.;

-(13) . Third Addendum to Lthe Technical Supplement to the Petition to Increase Power Level, Nine Mile Point Nuclear Station, Docket 50-220, December 1970.

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