ML19344A342

Monthly Operating Rept for Jan 1971

ML19344A342
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Site:	Midland
Issue date:	02/12/1971
From:	IDAHO NUCLEAR CORP.
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ILAHO NUCLEAR COR"0 RATION Nuclear-Safety Program Division t.

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MONTHLY REPORT

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January 1971 The information contained in this report is preliminary and subject to further i

evaluation.

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TABLE'OF CONTENTS

- Page A. -

LOFT' Integral Experiments and Radiological Studies..

1

.B.

Loss-of-Coolant' Accident Analysis...-.

-2 C.-

-Technical Assistance in Reactor Safety! Analysis.........

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D... Separate Effects ~ Testing'. J.................:...... 10

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s Janutry 1971

' HIGHLIGHTS LOFT Integral Test Program Development: Drafts of both the LOFT

- Measurements - Requirements Document (MRD-1) and the Experimental Require-

-ments' Document (ERD-1) were transmitted to.the LOFT Project and the AEC-

- for' review.

Fifteen LOFT Project SDD's and one LOFT technical report were reviewed

~

and comments were submitted to. LOFT Project.

Loss-of-Coolant Accident Analysis: A sensitivity study for both the H0XY and THETAl-B codes indicates for MOXY that using a few radial nodes in. the-fuel.is not necessarily conservative as believed previously.

The structural analysis STRAP-D code (developed for the Nuclear Safety

' Program) has been used to calculate the response fo the ICPP facilities to the 1940 El Centro earthquake.

The FLOMOD code has been satisfactorily checked out on the IBM 360/75 computer. This code is based on the INC convectional-flow model and calculates behavior of fission products in a containment under natural response conditions.

Separate Effects' Testing: Semiscale Tests 847 and 848 involving emergency core cooling (ECC) were performed with the ECC injected directly -into the vessel inlet plenum. The initial conditions for the two tests were similar except for the initial flow (225 spm for Test' 847, 145_gpm for Test 848) and the core temperature difference (370F for

-Test 847 and 600F for Test 848).

For Test 848 a quick closing valve was shut at the end of ECC injection in order to prevent the ECC accumu ator P s from entering the vessel.

Preliminary analysis of the results of

..ese tests indicates little or no core cooling by the emergency coolant.

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- January 1971

'A.

LOFT Integral Test hrogram Development.

~

l.. : Highlights >

.Draf ts of both the - LOFT Measurements. Requirements Document '(MRD-1)

-and Experimental Requirements Document (ERD-1) were. transmitted to LOFT Project and AEC.

Fifteen. LOFT Project SDD's and one LOFT technical report were reviewed and comments were submitted to LOFT Project.

2.

Technical Activities.

A draft of th'e LOFT Measurements Requirements Document (MRD-1) was transmitted-to the AEC for review.

The base-line. document specifies the programmatic information-required from the LOFT integral tests.

Section 2 of the LOFT PSAR, " Experimental Program" was revised in accordance with an internal review and resubmitted to LOFT Project for.

inclusion in the LOFT PSAR.

The following fif teen LOFT Project System Design Descriptions were reviewed and conments s,ubmitted to the LOFT' Project Division.

CDD 1.1.5.2A Air-Cooled Condenser SDD 1.2.6 Containment Leak Test System SDD 1.3.6.A Plant, Instrument, and Breathing Air. System-

[

SDD 1.3.2.1A Television System

~

CDD 'l. l. 4. '2B Primary Coolant Pump l

CDD 1.1.4.1B

' Steam Generator.

CDD 1.1.4.4B Pressurizer L

SDD 1.2.10A

. Containment Vessel Pressure Boundary System-SDD 1.3.10A Electric Power Distribution System

- SDD 1.3.25 Radiological Grid System SDD 1.2.11A Containment Floor Membrane, Sump Liners, and Membrane Nozzle System SDD 1.3.7 Steam, Condensa6e, and Boiler Feedwater System SDD'1.3.24-Annunciator System

. FDD ' l.~ 4A -

Instruments'and Control System's SDD 1.4.2A-Data Acquisition and Visual Display Syst'em A-LOFI Technical Report,. " Preliminary MTA Dynamic Analysis," was also

. reviewed.

A~ draft of the' Experimental' Measurements Document (ERD-1) was transmitted' l

to the' AEC for ? review.. The base-line document' specifies [the proposed i

test sequence for the LOFT integral test program for the-first core.'

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i Janunry 2971-

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' Loss-of-Coolant Accident Analysis B.

11..

Highlights l-

.A sensitivity' study of the effect-of space step size for MOXY and

- THETAl-B indicates 'using a - few radial' nodes in ' the fuel is not necessarily conservative for MOXY as believed 'previously.

A similar-study must.

be made ~ to determine the ~ significance of the number of nodes for other heat transfer codes.

J The mathematical models for structural analysis of the Chemical.

= Processing' Plant-facilities have been complot,ed and use,d _ to calculate the response:of the' facilities-to the 1940 El Centro earthquake.

STRAP-D was modified to analyze the nonlinear response of a wall -

with soil against one face.-

The FLOMOD code, based on the INC convectional-flow model for behavior of fission products in a containment under natural response conditions,

~

has been satisfact,rily checked out on the IBM 360/75 computer.

2.

Technical Activities Core Thermal Model Development: Development continues on the single-rod,. single-channel code, THETAl-B. A procedure has been written for the computer to take the RELAP3 output (core power, flow, pressure, and inlet enthalpy) used in THETAl-B; and automatically prepare this output for input to_THETAl-B.

This. procedure will: eliminate manual preparation of the data thereby saving about three man days per blowdown transient analyzed.

In addition, an~ axial sensitivity study was made using THETAl-B.

The purpose of,the study was to determine how many axial levels are needed to adequately describe the cladding surface temperature and the~ fluid enthalpy during a depressurization computation.

Four different

~

- cases:were considered. For the first three cases considered 7, 11, and 19 increments..of _ equal rod length were employed with a maximum power factor of 1~.72.

For the fourth. case, the seven-axial-level representation and a maximum power factor of 1.70 were used.

Identical pressure... flow rate, inlet enthalpy, and normalized power curves were used as boundary conditions for all cases.

On the basis

of the limited number of cases considered, _ the conclusion was reached that 'as the nudber of axial nodes is increased the predicted maximum hot-spot cladding temperature _ increases also.

This study also indicates that an appreciable error in cladding surface temperature.and' fluid enthalpy may occur if only a few axial levels are used to' model the fuel, rod and channel.

The results of

'the study indicate _ that negligiblu difference exists between eleven

nodes and nineteen nodes.-

An additional conclusion reached was that J

J the peak axial: power factor should be' used -for the entire increment lLn whichlthe hot spot occurs.

If an' average value is'used for the

hot spot increment a' lower peak ' surface temperature will be- '.omputed.

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-January 1971-1 i

To.Iprovidelan independent check on the. calculational accuracy of MOXY, l

"results from MOXY were compared to'those from! HEAT-1 and THETA 1-B.

A l common: problem was run'with all three codes. Temperatures predicted by-1MOXY Jand HEAT-1'were virtuali,Cidentical,' the maximum discrepancy being &

0.070F atithe end of a four-seem.2 transient... Comparison of. temperature's predicted by:M0XY and THETAl-B at: the. end of.10 seconds.of blowdown '

showed:that 'as; the space. step, size.'and time step size approach zero, the'

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computed temperatures = converge. The.depet.'dence of cladding and fuel

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temperature on space step sizejcan'be seen in Figure-l.

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Figure;1 showsf that the error due to: space step size is negative g

. in, MOXY and positive in THETAl-B, and that the ~ temperature predicted by-MOXY.1s closer to the. temperature of. convergence. Thus, a given 4

accuracy can be. achieved with fewer nodes by using -hoXY than by using

.THETAl-B.'

An,impo. tant conclusion' that may; be drawn from this study is that -

' the use;of = tooifew fuel nodes in MOXY is :not necessarily conservative,

. e s. believed. previously. An analysis similar to the. one reported here must be~ made to. determine 'the significance of the number of nodes for.

~

other' heat transfer: codes.

r.

An empirical 1 film heat transfer. correlation has been developed that.

. predicts'the heat transfer coefficient in the core during, flooding ECC" for the' lower 6 feet'of-a112-foot core. This correlation is based on PWR-FLECHT. Group,I'and Group.II tests. The main improvement over previous j--

' work is ethat quenching is :now predicted. The parameters in this correlation-

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(1) time afteriflooding (seconds), axial elevation (feet), and flooding

'are:

rate -(inches per second). This. correlation is intended to be conservative; that is, it is.. intended to predict the' lowest film heat transfer coefficient-

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and the. longest quench time for selected PWR-FLECHT tests considered.

4 Q.

An zimprovement-in the correlation is now being;made that includes.

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-as a parameter the. ratio'of pover'at..a node to the~ maximum rod power.

' The improvcd correlation will be-based on -PWR-FLECHT experimental heat transfer coefficients at the _2,

4-, 6, 8-and 10-foot elevations and

' thus will ' apply over" the'. entire length 'of a.12-foot fuel-rod.

i Structural-Model Development: The mathematical"models for structural L

analysis of the following-Chemical Processing Plant facilities were j

. completed:

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'(1).E Spent-Fuel Receiving Facility ll CPP 6011-Main Process Building

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-January 1971 These models were used to calculste : the maximum stresses that would have.

been experienced by each facility if it had-been subjected to the 1940 El' Centro earthquake.

A-report on the. seismic analysis of the-Chemical

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Processing Plant ;is being' written.

For the dynamic analysis of the Spent Fuel Receiving Facility,- a wall with soil on one side had to be modeled.

Figure 2 shows the physical configuration and the model.. The response of tnis wall is nonlinear because of the' tendency of soil to slip along a.line A-A, shown in Figure l 2a, as the wall moves away from the soil. - Figure 2b shows the.

wall and the soil af ter a displacement of the wall and an associated slippage of the soil. -The wall, after soil slippage.has occurred, cannot return to its original position-because of the increased resistance of the soil.

Figure 2c shows an analogous model of the wall and the soil.

The'" ratchet" model_ is generated by modifying the forcing function for l

the structure in such a manner as to.cause the wall to experience a much greater resistance to motion in one direction than in the other direction.

A routine to condense out unneeded degrees-of-freedom has been added.

to-STRAP-MS and ch2ck problems were run to' determine the accuracy of the routine.'.The input-output routines of STRAP-MS are being modified for production usage..

L Containment Model Development:

Development of'a slip-flow vent model j

l for CONTEMPT-PS continued. The Von Neuman linear stability analysis method i

is being used to examine the stability of the one-dimensional. difference equations for slip ficw in' a constant cross-sectional area vent (2),

l Analytical Models Describing Fission Product Behavior: The FLOMOD

-program, which is based on the' Idaho Nuclear Corporation convectional-flow l

model fcr behavior of fission 1 products in a containment under natural response conditions,' has been checked out and is operating satisfactorily on the IBM 360/75 computer.

Steps re'maining'to complete this effort areito (1). introduce a small correction in the diffusivity expressions and' (2) document the program.

l The FPFM program, which describes the release of noble. gases from operating fuel pins and which-is used to determine an. upper limit for release of iodine-131 to the fuel-cladding gap, has'been compared with i

i i

(1)

Robert-D. Richtmyer and K. W. Morton, " Difference Methods for Initial-

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-Value Problems," John Wiley,- New York,' N. Y.

(1967).

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(2) ' Nuclear Saf ety Program Division Monthly-Report - November (Hai-436-70)

December.14,.1970..

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January 1971 additional fission gas.releasg) data obtained at higher. heat ratings than the data reported previously(

The model. adequately predicts gas release at' the higher heat ratings. The Idaho Nuclear Corporation

' steady-state release model (the basis for the FPFM program) can be

.used to obtain Equation 1, which relates release fraction, F, to fuel pellet radius, b, for. two fuel pellets of different radius but equal ~-

' heat rating.~

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Subscripts 'l and 2 refer to Fuel Pellets 1 and 2.

Equation 1 provided'

.the information in Figure 3 relating fission product release from the

- fuel-to the radius of the fuel pellet and to the linear heat rating,

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January 1971

' C.

Technical Assistance in Reactor Safety Analysis i-

- 1..

.Highligh ts l

None 2.

. Technical Activities

. Core Thermal Response Analysis: The decay heat sources considered

~by the RELAP3 code.wcre investigated.

RELAP3 computes a delayed fission source' through use of one group of prompt neutrons.and six groups of delayed neutron precursors. -

fit of the data of K. Shure(3) Fission product

• decay is computed from a curve

,for eleven gamma emitters.

Decay heating -

from U-238 capture products, U-239, and Np-239, is not considered by RELAP3.

.Some concern has been associated with the method of calculating the postblowdown core heat source for the' first 30 to 60 seconds of a loss-of-coolant accident. The principal energy source for thermal transients during this time is the stored energy contained in the fuel rods because of the steady-state. temperatures. Thus, decay heating from all three

~

.ources is a secondary energy source for the transients of principal concern, and the capture product heating is a minor contribution to this secondary energy source. Since the uncertainty in the fission product decay heating for short transient times' is larger than the capture product contribution, little additional error results from ignoring the capture product source. The current RELAP3 models approximate the decay heat source reasonably well, and the analyses performed using these ~ models' are valid for short ' term transients.

For long-term applications,' the fission product ~ decay heating becomes the dominant energy-source, and the capture product contribu-tion becomes'significant. llence, for long-term applications, an additional energy source from capture product decay should be included in this calculation.-

(3)

K..Shure,' " Fission Product Decay Energy," Bettis Technical Review Reactor Technology, WAPD-llT-24 (December 1961) pp 1-17.

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s Jcnuary 1971 D.

Separate Effects Testing 1.

Highlichts l

Semiscale Tests 847 aid 848 involving emergency core cooling (ECC) were performed with the ECC injected directly into the vessel inlet plenum.

Preliminary analysis of the results of these tests indicates little or no core cooling by the emergency coolant. The initial conditions for the two tests were similar except for the initial flow (225 gpm for Test 847,145 gpm for Test 848) and the core temperature difference (370F for Test 847 and 600F for Test 848).

For Test 848 a quick closing valve was installed in the line between the ECC accumulator and vessel inlet plenum.

The valve was closed at the end of ECC injection in order to prevent the ECC accumulator gas from e'ntering the vessel.

2.

Technical Highlights Single-Loop Semiscale: The third semiscale test involving ECC Test 847, was performed late in December with a high inlet break (4),

and with the system orificed to more closely typify the pressure drops of a large PRR. The break area-to-system-volume ratio was 0.0007 ft-1 The test was conducted with initial conditions of 2290 psig system pressure, 225 gpm primary f1 >id flow rate, 556 and-5930F fluid temperatures for the inlet aad outlet plenums, respectively, and a total core power of 1.1 MW.

The ECC accumulator fluid conditions were 610 psig and 1480F, ECC fluid was injected directly into the inlet plenum.

Core power was terminated 8.3 seconds af ter rupture.

Decompression was essentially complete 60 seconds after system rupture.

The fourth semiscale test involving ECC, Test 848, was performed with a high inlet break and with a quick closing valve installed in the ECC line. The break crea-to-system-volume ratio was 0.0007 f t-1 The test was conducted with initial conditions of 2280 psig system pressure, 145 gpm primary fluip flow rate, 543 and 6030F fluid temperatures for the inlet and outlet plenums, respectively, and a total core power of 1.2 }M.

The ECC accumulator fluid conditions were 610 psig and 1480F.

ECC fluid was injected directly into the inlet plenum.

Core power was terminated 6.5 seconds af ter rupture.

Deccmpression was essentially complete 60 seconds after system rupture.

Figures 4 through 7 present the vessel inlet nozzle pressure, the pressure differential between the ECC accumulator and the inlet nozzle, and the ECC flow rate for Test 847 and 848.

For Test 847 the pressure.

differential between the accumulator and inlet nozzle reached a maximum of 198 psi 24.5 seconds after rupture and dropped to about 163 psi at 38.5 seconds when ECC liquid injection was completed.

For Test 848 the pressure differential between the accumulator and inlet nozzle reached a maximum of 176 psi 25.7 seconds after rupture and dropped to about 160 psi at 38.3 seconds when' ECC liquid injection was completed, fae 2.6 ft3 of ECC--fluid. (sufficient to cover the core) was injected during 3

both tests.

The average ECC flow rate was 43 gpm (0.096 ft /sec) for both tests.

(4).

Nuclear Safety Program Monthly. Report for December ' (Hai-7-71),

January 5,1971.

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January 1971

. 'The path followed by the ECC fluid sybsequent to injection into-

the~ inlet plenum is reflected in the fluid density data shown in Figure 8 for Test 847 and Figure 9, for Test 848.

Densities in the inlet nozzle, inlet plenum, Land core had decreased to about '2 lb/f t3 by 9 seconds after rupture for both Tests - 847 an/ 848.. Immediately' af ter the start of ECC injection, the density at the -inlet nozzle began gradually increasing and the density in the inlet plenum decreased and remained at - about-l'lb/ft3 for at least '10 seconds after ECC initiation for both tests.

Evidence-thatlno significant amount of ECC fluid reached the core for.either. test.is exhibited by - the core fluid density which increased only 'slightly nat the start of ECC injection and decreased to essentially zero forithe remainder -of _ the decompression.

Similar behavior of ECC -

fluid was noted in' the first two-ECC tests. (Tests 845 and 846) reported p rc'vious ly (2,4),.

' Differences between the vessel and inlet plenum densities occurred

- during Tests 847 and 848.. The earlier increase in he inlet plenum

~ density for Test b48 is believed due to the lower in;tial system flow rate and the associated larger initial temperature dii 'erence across the. core, and the fact that the' power was reduced to zero 1.8 seconds -

earlier than for Test 847.- The earlier decrease in the vessel' inlet plenum and inlet nozzle densities for Test 847 (40 seconds for Test 847 and 47 seconds for Test 848) is believed caused by the ECC accumulator gas which increases the vessel pressure and forces the ECC liquid out

~

the ' inlet plenum and nozzle and subsequently out the break.

In Test 848 the ECC gas was valved off at the end. of ECC liquid injection (40 seconds).

The decrease in the inlet plenum fluid density indicated at 47 seconds for Test 848 in Figure 8 is caused by the increase in vessel pressure.

due to the generation of steam which forces the ECC liquid from the inlet plenum and out the break in the same manner as did the ECC accumulator

_ gas.

. Figures 10 and 11'show pin cladding temperature histories ' at. representa-l-

tive locations and elevations in the core'.

As shown in Figure 10, departure i

from nucleate boiling (DNB) occurred at the midplane of. the heated length l?

on Pin 62 (a central pin) 2.5 seconds af ter rupture but by 3.5 seconds

. etting 'of the pin ' cladding was reestablished.

Similar phenomena are w

u evidenced for the top of Pin'13 (an outer pin) for Test 848 as shown in Figure 11.

For both tests,.- DNB occurred at all locations within 5.6 L

seconds..af ter. rupture. - Subsequently, the ' pin cladding temperature increased at a rate. of 1500F/sec 'for -Test 84 7 and at a rate between 0

150 and 300 F/sec for Test 848 until core poser was shut off.'

I l

l Shortly af ter ECC injection was completed,- momentary cooling,was I

recorded at all pin thermocouple locations for' Test 847 and was reflected

'by decreases-in the inlet and: outlet plenum fluid temperatures as shown in Figure 12.

All~ temperatures recovered to temperatures near those

existing prior.to imomentary cooling. The cooling at 40.5 seconds is J

~

believedLcau' sed.by the passage through-the core of cool gas from the:

ECC. accumulator. JThis cooling. event also exists at 'about 40 seconds -

3 for Test !848 'as shown in Figures; 11 and.13, but its magnitude is

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! drastically reduced,' indicating:that the quick closing valve prevented"

'the accumulator gas from: entering the v'essel, s

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..u January 191:

The data presented for Semiscale Tests 847 and 848 indicate that the emergency coolant did not flood the core and had no significant cooling effect on the pin material temperatures.

In addition, after ECC injection was completed (about 40 seconds), the ECC liquid in the inlet plenum was forced out the break because of r.te increase in vessel pressure which was caused by the ECC accumulator gas (Test 847) or by steam generation (Test 848).

PWR-FLECHT: Analysis of BWR-FLECHT test data is complete for the high-pressure tests and is continuing for the spray tes:s.

Stainless steci samples and work authorization to perform emissivity measurements ha e been received by the Heat Transfer Division of TRW, Inc.

The emissivity data will be used to resolve current differences between the Ceneral Elcetric Company and Idaho Nuclear Corporation analyses of top-spray test results from Bundles SS-2 and SS-4.

Recent efforts to interpret data on high-temperature spray performance in BWR-FLECHT tests have included calculation of a simplified energy balance for the first 11 minutes of the Zr-4 test.

Analysis of calculational results shows a large amount of energy removed (about 50% of the electrical energy input) which cannot be accounted for, even though the amounts of metal-water reaction and steam superheating included in the energy balance were assumed larger than believed possible.

Apparently, significant errors exist in one or more of the measured test parameters (the electrical power input, the spray rate, or the bundle drain rate).

The analysis also shows conclusively that all water input to the bundle during the first minute of the transient was allowed to accumulate.

Thus, Test Zr-4 was actually a combination spray and flooding test rather than a spray-only test as had been intended.

PWR-FLECHT:

Analysis of the data from the co=pleted PWR-FLECHT tests is continuing.

The PWR-FLECHT test program was performed on the premise that data from the extensive (eighty-four experiments) stainless steel-clad bundle tests would be applicable for predicting the thermal performance of Zircaloy-clad reactor rods during ECC.

Recent efforts were made to confirm this applicability by comparison of the results of stainless steel and Zircaloy bundle tests.

Four Zircaloy-clad bundle tests were performed with initial tempera-tures of 2000, 2000, 2325, and 20000F and flooding rates of 10, 4, 6 in./sec j

(for 8 seconds, then 1 in./sec), and 1 in./sec, respectively.

Comparison I

of the average heat transfer coefficients for five rods at the bundle j

midplane of the four Zircaloy-clad bundle tests and of comparabic stainlers steel-clad bundle tests does not reveal consistent diffewences. Without the results-from the last two Zircaloy tests, differences in Zircaloy and stainless steel test data could be dismissed as being possibly 'the result of experimental uncertainty.

However, cooling was notably better during the third Zircaloy-clad test and poorer during the fourth Zircaloy-clad test than during comparable stainless steel-clad tests.

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January 1971 For the fourth Zircaloy bundle test a metal-water reaction energg l-contribution of two times -that obtained through use of. the. Baker-Just ecuation was necessary?to obtain calculations that agreed with the experi-atal data previously. reported (9).' Zirconium oxide' thickness measurements te -being made :on rods from the fourth Zircaloy t'est. ~The measurements of temperature versus time and the zirconium oxide thickness will yield a

~

. cross-check on - the magnitude of.. the metal-water reaction.

~ No film concl$sion' can presently be drawn regarding the differences between Zircaloy and stainless steel-clad bundle behavior. : Idaho Nuclear Corporation :and Westinghouse Electric Corporation personnel held' a meeting

- January 23 to' discuss.interpretatien of. the tosults.

4 l

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NS) ' : Louis' Baker, Jr., and l Louis C. Just, Studies of Metal-Water Reaction -

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'at.' High Temperatures, ANL-65<48 '(May 1962).

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