ML20030A514
| ML20030A514 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/31/1962 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090819 | |
| Download: ML20030A514 (17) | |
Text
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huum wl& C-/-bU CONSUFERS POWER COMPANY Y2 BIG ROCK POINT NUCLEAR PIid.T., 7'3.,
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AFENRENT NO.12
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Part I - Maximum Heat Flux at, Overpower -
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j Phase I Research and ~ Development s \\
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Program Part II - Increased Cor; Size (56 to 78 Bundles) - Phsse I Research and Development Program
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SI0l)TD8)9May31,1962
PART I Maximum Heat Flux At Overpower Phase I Research and Development Program 1.
Table 8.1 of the Final Hazards Summary Report (FHSR) 3 includes a maximum heat flux at overpower of 447 x 10 stu/hr-ft. This number, together with the stated maximum fuel center temperature at over-power of 4400 F, was intended to app'y only to the initial or original fuel prior to the addition of Phase I otvelopmental fuel to the core.
By intention, no maxinum heat flux was specified in the FHSR for either the Phase I developmen:al fuel described in Section 10 thereof or the original fuel when mixed with developmental fuel, because more limiting criteria were established.
Subsequently, in the Proposed Technical Specifications, 3
dated January 15, 1962, a maximum heat flux at overpower of 540 x 10 Btu /hr-ft was specified (in Section 5.4.l(b)) as one of the principal core operating limitations. This heat flux was determined from Phase I research and development fuel design features presented in Section 5 3 1 of the Proposed Technical Specifications.
In Section 5 2.1 of the Proposed Technica: Specifications the principal calculated thermal and hydraulic characteristics of the 3
original core included the 447 x 10 Btu /hr-ft maximum heat flux at overpower (127%157Mwt). However, these characteristics were not in-tended to be specific operating limitations; as previously mentioned, operating limitations were specified in Section 5.4 of the Proposed Technical Specifications. This is illustrated in Section 5 3 2 of the Proposed Technical Specifications, which presented data, including a particular maximum heat flux at overpower, for a representative test to be conducted during Phase I of the developm&nt program.
In this example 3
the maximum heat flux at overpower was calculated to be 475 x 10 Btu /hr-ft.
3 The limit of 540 x 10 Btu /hr-ft was established before the specific dimensions of the Phase I developmental fuel had been fixed.
The exact rod diameter is now fixed at 0.425 inches and 3
e the associated maximum heat flux at overpower is 510 x 10 Btu /hr-ft 3
rather than 540 x 10 Btu /hr-ft.
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Tne following data are for a representative test to be conducted during Phase I of the development program. -The limiting i
parameter in this case is the 510,000 Btu /hr-ft heat flux, with reactor g
power below the specified operating limit (157 Mvt).
This test will be conducted only to the extent tbtt the stability criteria of measured zero to peak flux amplitude is less than 20 per cent. The anticipated worst conditions with regard to stability have been shown by calculation to have substantial phase margins, indicating that the core operating limita-tions will not be exceeded.' Other Phase I development tests, though LAffering in specific variables which may be limiting, will comply with the operating limitations specified in Section 3 Reactor pressure, psia 1,050
}
Core site, including 8 development full bundles, number 56 Reactor thermal power, Mvt 157 Average core heat flux, Btu /hr-ft 124,000 Average core power density, kv/l h5 m
Overpower pes}.ng factor 1.27 i
R&D Original Fuel Fuel Burnout ratio, minimum at overpower 19 19 5
Maximum heat flux, Btu /hr-ft, at overpower 510,000 h60,000 Maximum fuel rod power, kv/ft, at overpower 16.6 13 6 j
Maximum fuel center temperature, F, at overpower 3,750 3,350 Maximum cladding temperature, F, at overpower 595 625 j
3 The test program shall include the general areas of de-velopnental fuel irradiation, reactor core performance testing, stability and transient testing, and power distribution and physics testing. The limits imposed on these tests shall ~oe the most restrictive of those described below:
(a) The burnout ratio shall be shown by calculation to be at least 1
1 5 for all reactor operation test conditions. The burnout
- ratio is defined as the minimum ratio of the design burnout limit heat flux to the reactor heat flux as calculat ed using
.I
" Burnout Lianit Curves for Boiling Water Reactors," by E. Jansseu and S. Levy APED-3892.
h.
3 3
appropriate allowances for flux distributions, flow distriba-tions, effects of power transients and uncertainties.
(b) Develspment fuel clad stress shall not exceed 90% of yield strength calculated for end of life a+. 125% of rated heat flux, and zero reactor pressure.
(c) The heat flux shall not exceea 510,0t. Btu /hr-ft, and the fuel rod power generation per unit length
'all not exceed 16.6 kv/ft.
These quantities shall be determined 1 tse of the same allowances 4
described in the burnout ratio calcult is.
(d) Test conditions shall be such that the r sured steady state zero to peak flux amplitudes shall not exceed _ i of the average operat-ing flux level for a given test.
In addit 7,
the zero to peak flux amplitude shall not exceed 20% of the o erating level during any stability test which may be conducted.(1*
When rod oscillation is a part of the test, the initial amplitude t disturbance will be limited to one control rod notch or equivalent. When results of this limited controlled disturbance give evidence that the equiva-lent of multiple-rod or multiple notch oscillation will not exceed the stated flux criteria, these larger disturbances may be applied as part of the stability testing.
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Listed below is a summary of the operating limits which shall be observed with any core composed of both original and R&D fuel during the Fhase I tests.
Minimum burnout ratio, at overpower 15 Maximum Heat Flux, Btu /hr-ft, at overpower l
R&D Fuel 510,000 J
Original Fuel 460,000 MaximumFuelRodPower,Kw/Ft,atoverpower R&D Fuel 16.6 Original Fuel 13 6 Maximum fuel exposure, for a single bundle, Mwd /t 21,000 Maximum fuel clad stress, per cent of yield stress 90 4
i See Section 4
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Flux amplitude, zero to peak, per cent of operating level Steady state operation (noise) 20 Transient testing (noise and excitation) 20 Maximum steady state power level, Mwt 157 Maximum core power density, total core power divided by total core volume, hv/l 45 6
Minimumrecirculationflowrate,lb/hr 6 x 10 Reactor Operating Pressure, psia Minimum 800 Maximum 1,500 The only change between the original operating limits and
+ hose described herein involve the maximum heat flux and the maximum fuel rod power. Satisfactory experience at VBWR has been obtained at the proposed increased values, and shova that these limits are acceptable from the standpoint of fuel center temperature. Justification for this vill be found in Section h.
h.
An increased heat flux is reflected in many areas.
It affects the center temperature of the fuel, the fission gas release and associated clad stresses, and the burnout margin of operation.
The center temperature of a fuel element is obtained from the following equation.
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Equation No.1 T
=T c
y vhere T = Fuel center temperature - F T = Bulk water temperature oF y
Q/A = Heat flux - Btu /hr-sq-ft D = Fuel rod outside diameter - ft g
D = Fuel rod inside diameter - ft y
h = Nucleate boiling heat tgansfer coefficient at clad 0.D., (10,000 Btu /hr-ft - F) h = Heat transfer coefficient between fuel pellet and 8
C cladding, (1,000 Btu /hr-ft OF)
= %ema1 conducHy of fue1711% - h/MMO k y92 ey= Thermal conductivity of cladding material - Btu /hr-ft-F K
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5 Originally, calculations of the fuel center ttmperature were carried out by using a value of 1.1 to 1.15 Btu /hr-ft P for the effective thermal conductivity of UO. This value was based )n experi-2 mental data which gave void formation in the center of Dresden type fuel 2
at a heat flux of about 350,000 Btu /hr-ft.
It was assumed that the void formation was due to melting of UO and correcponded to a center fuel 2
temperature of 4950 to 5000 F.
If this original method of calculating 4
center temperature were to be employed for the propcsed limit the follow-ing temperatures vould result:
Heat Flux Btu /hr-ft Center Line Temperature F
447 x 103 4h00 3
510 x 10 5150 on this basis, the proposed increased heat flux limit means operation at 200 F above void formation.
It should be pointed out that when the central void formati'>n limit was originally adopted, it was al-ready recognized that no proof existed that operation beyond this limit was detrimental. The void formation limit was always considered to be a
" soft" limit in the sense that operation beyond it had not been chown to produce any hazards and for that reason operating limits were originally proposed which did not specify a limiting heat flux or center fuel tempera-ture.
Operation beyond void formation has since been justified, as follova:
(a) Laboratory experimenta have shown that void formation occurs at temperatures of 3000 F instead of 4950 F as originully assumed.
(b) The UO thermal conductivity value of 1.1 to 1.15 has been found 2
to be low.
This is understandable in view of the reduced tempera-ture at which void formation occurs.
(c) Experiments have been carried out at conditions beyond those lead-ing to central void formation and they have led to satisfactory fuel operation.
I (1) Nuclear Metallurgy, Vol 6, AIME, November 1959 by J. L. Lates and W. E. Roake Hanford Reports IN59575, IN57113, IN60828
0 I
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Based upon this information, a more realistic model of pre-dicting center temperature has.been developed.
This model is still postu-lated upon Equation No. 1, except that it utilizes an equation proposed by F
Bates for the calculation of the therm a conductivity of UO. Bates has 2
proposed the following expression.
-2 3
h 2 55 x 10 T
Equation No. 2 K
=
where K =Thermalconductivitywatts/cmK T = Abcolute temperature K i
t P = Specimen density Po = Theoretical crystallographic density
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Application of the Bates correlation to the original Dresden central void formation tests gives a center fuel temperature of 3300 F i
which agrees with the p"edicted temperature for this type of phenomenon.
Application of Bates correlation to the original proposed heat flux limits t
3 gives tbs following center temperature:
1 Orig 1.'el Limit Present Limit Heatfluxh47,000 Btu /hr-ft 510,000 Btu /hr-ft I
Center temperature 3300 F 3800 F
'The above calculations show that with the proposed limit the fuel operates at 1100 F below fuel melting and about 500 F above void forma-tion. This condition is not expected to cause any safeguard proble.ns. This is demonstrated by the following tabulation of maximum heat flux and red power for various types of prototype fuel successfully irradiated in VB'n'R.
VB'n'R Testing 4
Maximum Heat Flux Maximum Rod Description Btu /hr-ft Power Kv/ft 3
Dresden Developnent 525 x 10 23 i
l Savannah hC8 x 103 17,g Consumers HPD Fuel (1 527 x 10 16.2 3
(1)HPD Development Project - Monthly Progress Letter No. 27 for April 1962,
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= dated May 1, 1962.
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i' It is particularly important to note that developmental j'
assemblies have been irradiated under the developmental program associated with the Big Rock Point at rod power values corresponding to these of the proposed limit. The Dresden development tests have been performed well in excess of the values, while the Savannah fuel assemblies have attained burnup in excess of 2000 Mvd/T at a rod power in excess of that requested.
i Under the Euratom Program several capsules have been i
irradiated at average surface heat fluxes of 560,000 and 595,000 Btu /hr-ft.
Respective peak heat fluxes were 960,000 and 860,000 Btu /hr-ft.
No detri-mental effects were noted.
Calculations were made using the Bates correla-tion to predict the temperature distribution within the fuel.
Destructive examinttion of the capsules revealed excellent correlation between calculated heat flux to produce center melting and actual heat flux that produced center melting.
Under the Fuel Cycle Program, fuel assemblies are being irradiated in the VBWR under conditions of center UO melting.
Calibration 2
runs were performed to determine the heat flux at which incipient melting occurred. A fuel rod containing 1.235" diameter pellets was operated in the BVWR with a maximum surface heat flux of 535,000 Btu /hr-ft, correspond-ingto55kv/ft. No detrimental effect was noted. Results of destructive examination shoved that center melting had occurred over a length of 14 inches. The observed microstructure and center void formation found upon i
destructive examination was ccusistent with the temperatures as calcula+ed using the Bates data.
Considerable irradiation and examination of UO f"*1 18 2
being performed at the Hanford Atomic Products vperation. Results of center melting experiments are repo:-ted in HW-73o72, " Irradiation of vo '"
2 W. E. Roake, March 1962.
l Irradiation of sintered, svaged, and vibration compacted l
UO at surface heat fluxes in the range of 600,000 to 900,0C0 Btu /hr-ft 2
were performed to obtain central melting. Results of destructive examina-tion showed that center melting did occur and was consistent with the Bates data. Maximum heat rating for the sintered, svaged, and vibration
~
8 compacted UO samples were 31 kw/ft, 26 kv/ft, and 39 hv/ft, respectively.
2 The report presents additional experimental data on irradiated svaged, vibration compacted, and sintered UO to support the original thermal con-2 ductivity recults published by Bates.
'J'he effects of increased heat flux were calculated in GEAP-3851. Clad stresses based upon a maximum heat flux ol' 550,000 Btu /hr-ft are shown in Table B-2 of that report.
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i PART 2 r
Increased Core Size (56 to 78 Bundles)
Phase I Research and Development Program i
l.
The initial or original core vill consist of 56 fuel
-bundles or assemblies, as described in the FHSR and the Proposed Technical Specifications, dated January 15, 1962.
Phase I of the Research and De-velopent Program includes four instrumented and eight develop:nental fuel i
bundles that will replace twelve original core bundles in the 56 bundle j
core. Figure 10.1 of the FHSR provides an approximate arrangement of the 56 bundle, 157 Mvt core with Phase I testing.
2.
Section 10.2.2.13 of the FHSR summarizes the intent to
[
conduct a sequence of tests with an enlarged core (78 bundles), during f-Phase I of the development program. Figure 7.1 on page 3 is a typical i
layout for the 78 bundle core with 32 control rods.
3 It is important to reconfirm at.the outset that the requested 78 fuel assembly core operation ir to be limited to 157 Mvt.
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Under this condition, the maximum fuel her.t flux and center temperatures i
are well below those associated with the 56 fuel assembly core.
4.
The stuck rod criteria vill be observed in any attempt 1
i to load the larger core size. At the time the power demonstration tests have been completed and the Phase I development tests with the 56 bundle core have been completed, it is expected that fuel burnup will have de -
creased the core reactivity sufficiently to permit successful loading l
of the 78 bundles.
5 The effects of the increased core size, in terms of tr a sient analyses previously reported (Section 12, FHSR), are as i
follows:
(a) Steam Bypass Valve Testing The transient caused by testing the bypass valve is of such short duration and small magnitude that it has a negligible effect on reactor pressure and flux. Fuel temperature and surface. heat flux are not affected appreciably. Thus there vill be little effect from increased core size.
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2 (b) Pressure Regulator Set-Point Changing
+
The result of increasing the initial pressure regulator set point is a momentary reactor isolation until pressure rises to the new set point. As a result, flux peaks to a level approach--
ing the scram setting for pressure changes at the maxbnxn rate.
In the limiting cases, either bypass valve action or scram, the maximum expected change in surface heat flux is approximately 12%. For a 78 bundle core, the heat flux is reduced and a 12%
increase causes no concern.
(c) Control Rod Withdrawal and Insertion I
Continuous control rod withdrawal results in scram for a re-l activity addition in excess of $1.00.
For the 157 Mwt core, the maximum rise in fuel temperature and surface heat flux is approximately 120%. For the increased core sizes, the same relative increase in heat flux and fuel temperature is obtained and this poses no problem due to the reduced heat flux and fuel temperatures that exist at increased core size.
(d) Rod Drop Accident The rod drop-out accident hea been analyzed for the maximum rod worth config tration obtained in a core containing 84 fuel assem-blies. Operation with a 78 fuel assembly core should lead to less severe consequences as the worst configuration hse already been analyzed.
(e) Loss of Recirculating Pumps This analysis has been done for the 56 bundle core at 157 Mut and at 1050 psia and 1500 psia. It has also been done for an 84 fuel bundle core at 240 Mwt and 1500 psia.
Comparison of i
the results at 1500 psia with a 56 fuel bundle and the 84 fuel bundle core show that the burnout ratio with the loss of pumps-is higher for the Bh fuel bundle core in spite of its increased initial output. Further, at 1050 psia and 157 Mwt the 56 fuel i
bundle core including developmental fuel assemblies has a burn-out ratio of 1 9 at overpower conditions.
The 84 fuel bundle
. core at the same conditions has a burnout ratio of 2 7 The 78 fuel bundle core at the same conditions is expected to have i
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-dw FIGURE 2.1 TYPICAL LAYOUT FOR 78 BUNDLE CORE
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a burnout ratio of 2.6.
The loss of recirculating pumps vill,- therefore, be less severe in the 78 fuel bundle core than in the 56' fuel bundle core because the 78 fuel bundle core has a higher burnout ratio before the loss of pumps occur.
(f) Loss of Electric Load The loss of electric load is a very fast transient. This has been analyzed for 157 Mvt and 2ho Mvt cores.
In the limit-in;; case, which is a stop valve closure, the reactor might scram, sithough this is not expected.
Bypass valve action is expected to limit'the flux to less than 120% of normal for all other load rejections.
Fuel. temperature and thermal heat flux are expected to remain below 110% normal in any case.
Thus, with the reduced heat fluxes of the 78 bundle core, the consequences are less serious than with the 56 bundle core.
(g) Closure of Steam Line Back-Up Isolation Valve The reactor vill be scrammed on isolation valve closure.
This valve closes comparatively slowly, and scram is initiated at about h0%. stroke.
The resulting transient pesks vill be lower than if scram were initiated by high flux. Here again, the reduced initial heat fluxes should reduce the consequences of the acci-i dent in a 78 bundle core.
(h) Coincident Steam Shutoff With Failure to Scram Safety valve settings were made in accordance with the criteria for maintaining vessel integrity. In the 78 bundle core, the burnout ratio will be greater, than in the 56 bundle core, as previously noted, and the possibility of burnout vill be re-duced in the_ event of a safety valve incident.
f (1) Loss of Condenser Vacuum The stop valves are closed and the reactor is scrammed simul-taneously on loss of condenser vacuum. Flux, fuel temperature, and surface heat flux peaks are negligible for this case. The increased core size should reduce the effects of this accident.
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(-j ) Riser Rupture, Main Steam Line Break 2
These are maximum credible accidents and have been analyzed and reported for the vorst expected case.
These cover the tse of i
the 78 bundle' core.
(k) Start-Up Accident This accident has been analyzed for the maximum rod worth con-figuration obtained in an 84 fuel bundle core. This analysis, therefore, covers the case of a 78 fuel bundle core.
(1) Cold Water Accident The cold water accident is a very slow transient in which fuel temperature and surface heat flux follow neutron flux very closely.
The accident is terminatec by a flux serem.
Tf the flux scram should fail to occur, the pressure scram occurs at 50 psi above operating pressure.
center fuel temperature at l
the hot spot at this time is high enough that some melting I
might occur. However, the fuel does not burn out. For the 78 bundle core at 157 Mwt, the center temperature vill be re-duced and hence, this core vill be further from the burnout condition.
I (m) Fuel Lotding Here again, the condition assumed for the analysis was max-1 imized and covers the case of a 78 bundle core.
l 6.
The design of the Phase I developmental fuel is such as to give nuclear characteristics equiva'ent to the original core. The coefficients based on leakage from a minimum critical array of this type fuel are:
Void coefficient (68 F, zero voids) Ak/% voids.
-0.2
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i Temperature coefficient (68 F, zero voids) 6 k/ F
+0.h x 10
-The coefficients become more negative with increasing temperature. The developmental fuel'is more reactive than the original fuel loading by 2 5%Ak/k. The effect'of samarium poisoning in the original fuel com-c bined with this higher reactivity developmental fuel is calculated to-provide the same reactivity as the clean original core. Shutdown margin
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'and reactivity balance.also remain.the same as in the original core.
6 7
The flux smplitude limits chosen do not of themselves represent firm limits. The scram limit is determined by the burnout heat flux and the fuel performance design values for steady state operation.
The limit of 20% for zero to peak flux amplitude variation is chosen to give reasonable tasurance that the reactor vill not be scrammed by opera-tion at the test conditions.
There are no known harmful effects from operation vi.th flux. The VEWR vas operated with zero to peak flux varia-tions of 15$' to 20% during flux noise and rod oscillation testing in November and December, 1961, with no harmful effects or operating diffi-culty. One scram occurred when a flux oscillation reached 25% above the operating power level, i
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