ML20098F535
| ML20098F535 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 09/26/1984 |
| From: | Tramm T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 9225N, NUDOCS 8410030097 | |
| Download: ML20098F535 (13) | |
Text
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)_ one First National Plaz1. Chicago, Illinois
'd
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) Address Reply ty P:st Ottice Rox 767
\\
f Chicago, Illinois 60690 September 26, 1984 7 -
Mr.-Harold R. Denton, Director Office of Nuclear Reactor ~ Regulation U.S. Nuclear Regulatory Commission Washington,'DC. 20555 Sub ject:
Byron Generating Station' Units 1 and 2 Braidwood Generating Station Units 1 and 2 Reactor Coolant Pump Transients NRC Docket Nos.-50-454/455 and 50-456/457 References.(a):
June 7, 1982 letter from T. R.
Tramm to H. R. Denton.
/
(b):
May 2, 1984 letter from T. R. Tramm to H. R. Denton.
Dear Mr. Denton:
.This letter provides additional information regarding postulated
. reactor coolant pump locked rotor and shaft break transients for the Byron /Braidwood units.
NRC review of this information should close Confirmatory-Issue 30 of the Byron SER.
In reference (b), Commonwealth Edison provided advance copies of
-revised FSAR pages which document a recent reanalysis of the locked rotor transient.
After review of these pages by the NRC, it has become
= apparent that other FSAR changes were also necessary.
Enclosed are advance copies of revised FSAR section 15.3.3 and
. Tables'15.0-9 through 15.0-12.
These changes will be incorporated into the FSAR at the earliest opportunity.
These changes make it clear that, although no fuel rods experience DNB, the analysis conservatively assumed iDNB at the core hot spot.in determining the upper limits of clad tempera-ture and zirconium-water reaction.
Please address further questions regarding this matter to this office.
One signed original and fifteen copies of this letter and the attachments are provided for NRC review.
Very truly yours, g [g %
A 0
T. R. Tramm Nuclear Licensing Admininstrator 1m Attachment goo l
t j
9225N
y
. v
-s, i
B/B-FSAR TABLE 15.0-9 SECONDARY COOLANT EQUILIBRIUM IODINE ACTIVITY BASED ON 0.1 uCi/gm OF g,
DOSE EQUIVALENT I-131 CONCENTRATION ISOTOPE (gCi/gm)
I-131-.
0.066 I-132 0.239 I-133 0.106 I-134 0.016 I-135 0.058
~
f Ebs
\\
15.0-41 h
+e-e
y B/B-FSAR-TABLE 15.0-10 REACTOR COOLANT IODINE ACTIVITY BASED O';
60 pCi/gm OF DOSE EQUIVALENT I-131 AND REACTOR COOLANT NOBLE GAS Im7ENTORY BASED ON 1% FUEL DEFECTS ACTIVITY ISOTOPE (pCi/gm)
'l I-131 39.76
'I- '.3 2 14.31 I-133 63.62 I-134 9.54 I-135 34.99 Xe-133 398.03 Xe-133m 4.39 Xe-135 8.92
-Xe-135m 0.183 Xe-138 0.992 Kr-85 12.46 Kr-85m 2.97
.Kr-87 1.70 Kr-88 5.24 h..
s I
15.0-42 L.
'~
i
- }
TABLE 15.0 POTENTIAL OFFSITE DOSES DUE TO ACCIDENTS
- BYRON STATION l
DOSE-(2 HOURS)'AT EXCLUSION DOSE (COURSE OF ACCIDENT) AT AREA BOUNDARY LOW POPULATION ZONE (445 meters).
(4848 meters)
FSAR 1
Postulated Accident Section Thyroid (rem) Whole Body (rem) Thyroid (rem)' Whole" Body (rem)
Steam Line Break 15.1.5 Conservative 2.89(+1)
- 4. 29 (-1) 4.22 (+0) 2. 81 (-2 )
Realistic 1.23(-5)
- 5. 30 (-8 )
6.14(-7) 3.22(-9)
M Locked Rotor 15.3.3 s.
o Conservative
- 9. 9 5 (-2) 7.20(-4)
- 1. 84 (-2)
- 9. 6 3 (-5)
I T
Realistic 2.98(-5)
-6. 09 (-8) 1.98(-6)
- 3. 90 (-9) y a
5 j
Rod Ejeccion 15.4.8 Conservative
- 4. 4 3 (+1) 4.19 (-1)
- 3. 90 (+0) 1.66(-2)
Realistic 1.97(-5) 3.56 (-8)
- 3. 75 (-6)
- 5. 0 0 (-9 )
' Steam Generator 15.6.3 Tube Rupture Conservative 2.16 (+1) 2.62 (-1)
- 8. 26 (-1) 8.12 (-3 )
Realistic
- 9. 25 (-7)
- 6. 21 (-4 )
2.16 (-8) 1.45(-5)
LOCA 15.6.5 Containment Leak Conservative 1.14 (+2)
- 5. 26 (+0)
- 1. 58 (+1)
- 3. 4 2 (-1)
Realistic 4.04 (-6)
- 9. 2 3 (-8 )
- 4. 4 2 (-7) 6.85(-9)
~
ESP Equip. Leakage Conservative
- 7. 7 5 (-1)
- 2. 02 (-3)
- 1. 62 (-1)
- 1. 8 7 (-4 )
Realistic
- 3. 02 (-8) 8.66(-11)
- 1. 4 7 (-8) 1.49(-11) i e
j
g.
TABLE-15.0-12
-POTENTIAL OFFSITE DOSES DUE TO ACCIDENTS *
~
BRAIDWOOD STATION DOSE (2 HOURS) AT EXCLUSION DOSE (COURSE OF ACCIDENT). AT -
AREA BOUNDARY LOW POPULATION ZONE (485 meters)
(1811 meters)
FSAR Postulated Accident Section Thyroid (rem)' Whole ' Body (rem)' Thyroid (rem) Whole Body (rem).
Steam Line Break 15.1.5 Conservative 3.90(+1) 5.79(-1)
- 1. 76 (+1) 1.17 (-1)
Realistic 1.35(-5)
- 5. 7 9 (-8)
- 2. 70 (-6) 1.42(-8)
W Locked Rotor 15.3.3 g.
7' Conservative
- 1. 3 4 (-1)
- 9. 7 2 (-4)
- 7. 69 (-2)
- 4. 03 (-4) l
- o Realistic 3.25(-5) 6.66 (-8) 8.73(-6) 1.72(-8) g a
n Rod Ejection 15.4.8 s
m Conservative-
- 5. 99 (+1) 5.66 (-1)
- 1. 79 (+1) 7.07(-2)
Realistic 2.16 (-5) 3.89(-8)
- 1. 8 4 (-5) 2.39(-8)
Steam Generator Tube Rupture 15.6.3
- 3. 54 (-1)
- 3. 4 5 (+0)_
- 3. 39 (-3)
Conservative 2.91(+1)
Realistic
- 1. 01 (-6 )
6.78 (-4)
- 9. 50 (-8) 6.37(-5)
LOCA 15.6.5 Containment Leak Conservative
- 1. 54 (+2) 7.10 (+0 )
- 7. 32 (+1)
- 1. 4 6 (+0)
Realistic 4.41(-6)
- 1. 01 (-7 )
2.16(-6)
- 3. 24 (-8 )
ESF Equip. Leakage Conservative
- 1. 05 (+0 )
2:. 7 3 (-3)
- 7. 91 (-1)
- 8. 44 (-4 )
Realistic
- 3. 30 (-8) 9.46(-11) 8.02(-8) 7.28(-11) l
e
(
B/B-FSAR 1 Plant characteristics and initial conditions are further dis-cussed _in Subsection 15.0.3.
With three loops operating, the maximumLpower level (including errors) allowed in that mode of Loperation is assumed.
'For.the peak pressure evaluation, the initial pressure is con-
~
servatively estimated.as 30 psi above nominal' pressure (2250
, psia)'to; allow for errors in the pressurizer pressure measure-ment and control channels.
This is done to obtain the highect possible rise in the coolant pressure during the transient.
To obtain the maximum pressure in the primary side, conservatively high loop pressure drops are added.to the calculated pressur-izer pressure.
The pressure responses shown in Figures 15.3-18
<and 15.3-22 are=the responses at the point in the reactor coolant system having the maximum pressure.
Foria conservative analysis of thermal behavicr, the hot spct
< evaluation-assumes that DNB occurs at the initiation of the transient'and continues throughout the transient.
Although no rods-are predicted to be in DNB, this assumption reduces heat transfer. to the coolant and results in conservatively high hot spot temperatures.
' Evaluation of the Pressure Transient After pump seizure, the neutron flux is rapidly reduced by control' rod insertion.
Rod motion-is assumed to begin one second af ter the flow in the -af f ected loop' reached 87% of nomi-nal flow.
No credit is taken for'the pressure reducing effect of _ the -pressurizer-relief valves, pressurizer spray,. steam dump Lor. controlled feedwater flow after plant trip.
Alth'ough these operations are expected to occur and would result in a lower peak' pressure, an' additional degree of con-
-servatism is provided by ignoring their effect.
The pressurizer safety valves are full open at 2575 psia and their capacity for steam relief is as described in Section 5.4.
..e Evaluation of Hot Spot Temperature in the Core During the Accident
.Although no rods are predicted to be in DNB, for conservatism in for this accident, DNB is assumed to occur at the hot spot and therefore, an evaluation of the consequences with
-the core, Results respect to fuel _ rod thermal transients is performed.-
obtained'from analysis of this " hot spot" condition represent the upper limit with respect to clad temperature and zirconium In the evaluation, the rod power at the hot
. water reaction.
spot is' assumed to be 3.0 times the average rod power (i.e.,
3.0) at the initial core power level.
F
=
g 15.3-8
e: -
T B/B-FSAR i
c Film Boiling Coefficient
.The' film; boiling' coefficient is calculated in the PACTRAN Code
' using.theLBishop-Sandberg-Tong film boiling correlation.
The fluid properties are evaluated at film temperature (average between wall and bulk temperatures).
The program calculates l'
the film. coefficient at every-time step based upon the actual heat transfer cor.61tions at' the time.
The neutron flux, system pressure,: bulk density and mass flow rate as a function of time are used as program input.
I1 i
4.
15.3-8a c.
_ ~. -.... _
O B/B-FSAR For this analysis, the initial values of the pressure and the bulk density are used throughout the transient since they are the most conservative with respect to clad temperature response.
For concorvatism, DNB was assumed to start at the beginning of the accident, although as noted above, no rods were predicted to be in JN3.
Fuel Clad Gap Coefficient The magnitude and time dependence of the heat transf er coef fi-cient between fuel and clad (gap coefficient) has a pronounced influence on the thermal results.
The larger the value of the gap coefficient, the more heat is transferred between pellet and clad.
Based on investigations on the effect of the gap coefficient upon the maximum clad temperature during the tran-sient, the gap coefficient was assumed to increase from a steady state valge consistent with initial fuel temperature to the initiation of the transient.
10,000 Btu /hr-ft. cF at Thus, the large amount of energy stored in the fuel because of the small initial'value is released to the clad at the initia-tion of the tran3ient.
Zirconium Steam Reaction The zirconium-steam reaction can become significant above 1800*F (clad temperature).
The Baker-Just parabolic rate equation shown below is used to define the rate of the zir-conium steam reaction.
2
~45,500 -
d (w )
6 33.3 x 10 exp
=
dt.
.l.986 T.
where:
r w = amount reacted, mg/cm2 t = time, sec T = temperature, 'F The reaction heat is 1510 cal /gm.
The effect 'of zirconium-steam reaction is included in the cal-culation of the " hot spot" clad temperature transient.
Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Subsection 15.0.8 and listed in Table 15.0-7.
No single active failure in any of these systems or equipment will adversely affect the conse-qucnces of the accident.
15.3-9
r b
k D/B-FSAR Results Locked Rotor with Four Loops Operating
' Transient results for this case are shown in Figures 15.3-17 through 15.3-20.
The results of these calculations are also summarized in Table 15.3-2.
The peak reactor coolant system pressure reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits.
Also, the peak clad surface temperature is considerably less than 2700 F.
It should be noted that although no rods were predicted to be in DNB for this analysis, the clad temperature was conservatively calculated assuming that DNB occurs at the initiation of the transient.
Locked Rotor with Three Loops Operating Bounding transient results for this case are shown in Figures 15.3-21 through 15.3-24.
The results of these calculations are also summarized in Table 15.3-2.
The peak reactor coolant system pressure is slightly higher than for the previous case, but is still less chan that which would cause stresses to exceed the faulted condition stress limits.
The clad temperature transient is less severe than for the previous case.
The calculated sequence of events for the two cases analyzed is shown in Table 15.3-1.
Figures 15.3-17 and 15.3-21 show the core flow rapidly reaches a new equilibrium value.
With the reactor -tr ipped, a stable plant condition will eventually be attained.
Normal plant shutdown may then proceed.
15.3.3.3 Radiological Consequences The evaluation of the radiological consequences of a postulated seizure of a reactor coolant pump rotor (Locked Rotor Accident-LRA) assumes that the reactor has been operating with a small percent of defective fuel and leaking steam generator tubes for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant.
As a result of the accident, radionuclides carried by the primary coolant to the steam generators, via the leaking tubes, are released to the environment via the steam line safety or power operated' relief valves.
The major assumptions and parameters used in the analysis are itemized in Table 15.3-3.
.\\
15.3-10 L'
l D/B-FSAR 15.3.3.3.1 Source Term The concentration of nuclides in the primary and secondary system prior to and following the accident are determined as follows:
a.
The iodine concentrations in the reactor coolant will be based upon preaccident and accident initiated iodine spikes.
1.
Accident Initiated Spike - The reactor trip associated with the LRA creates an iodine spike in the pri-mary system which increases the iodine release rate from the fuel to the primary coolant to a value 500 times greater than the release rate corresponding to the maximun equilibrium primary system iodine concentration of 1 Ci/gm of Dose Equivalent (D.E.) I-131.
The duration of the spike is assumed to be 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
2.
Preaccident Spike - A reactor transient has occurred prior to the LRA and has raised the primary coolant iodine concentration to 60 pCi/gm of Dose Equiva-lent I-131.
b.
The noble gas concentrations in the primary coolant are based on 1 percent defective fuel.
c.
.The secondary coolant activity is based on the D.E.
of 0.1 Ci/gm of I-131.
15.3.3.4 Conclusions a.
Since the peak reactor coolant system pressure reached during any of the transients is less than that which would cause stresses to exceed the faulted condition stress limits, the integrity of the primary coolant system is not endangered.
b..
Since the peak clad surface temperature calculated for the hot spot during the worst transient remains considerably less than 2700 F,
the core will remair in place and intact with no loss of core cooling capability.
Note that for conservatism, this evaluation assumed DNB to occur at the initiation of the transient and continues throughout the transient, although no rods were predicted to be in DNB.
\\
15.3-11
e
+
B/B-FSAR c.
The radioactivity released to the environment as the result of a postulated LRA is presented in Table 15.3-4.
.The resulting thyroid and whole body doses at the exclusion area boundary and at the low-population zone outer boundary are presented in Tables 15.0-11 and 15.0-12.
15.3.3.5 Locked Rotor With a Concurrent Power Ooerated Relief Valve (PORV) Failure A locked rotor event with a concurrent PORV failure was also evaluated.
In evaluating the radiological consequences of this event,. water level in the aff ected steam generator is assumed to be lost and hence, no credit for iodine partitioning is taken.
The consequences for this event are bounded by the steam line break consequences presented in Section 15.1.5.
15.3.4 Reactor Coolant Pump Shaft Break 15.3.4.1 Identification of Causes and Accident Description The accident is postulated as an instantaneous failure of a reactor coolant pump shaft, such as discussed in Section 5.4.
Flow through the affected reactor coolant loop is rapidly reduced, though the initial rate of reduction of coolant flow is greater for the reactor coolant pump rotor seizure event.
Reactor trip is initiated on a low flow signal in the affected loop.
Following initiation of the reactor trip, heat stored in the fuel rods continues to be transferred to the coolant causing the coolant to expand.
At the same time, heat transfer to the shell side of.the steam generators is reduced, first because the reduced flow results in a decreased tube side film coefficient and then because.the reactor coolant in the' tubes cools down while the ~shell side temperature increases (turbine steam flow is reduced.to zero upon plant trip).
The rapid ' expansion of the coolant in the reactor core and reduced heat transfer in the steam generators cause an insurge into the pressurizer and - a pressure increase throughout tbc reactor coolant system.
The insurge into the pressurizer compresses the steam volume, actuates the automatic spray system,' opens the power-operated relief valves, and opens the pressurizer safety valves, in that sequence.
The two power-operated relief valves are designed for reliable operation and would be expected to funccion properly during the. accident.
However, for conservatism, their pressure reducing Effect as well as the pressure reducing effect of the spray is not included in the analysis.
15.3-11a
~-
B/B-FSAR TABLE 15.3-3 ASSUMPTIONS USED FOR THE LOCKED ROTOR ACCIDENT EXPECTED DESIGN Power 3565 MWt 3565 MWt l
Fraction of Fuel with 0.0012*
See Defects Subsection 15.3.3.3 Reactor Coolant Activity ANS-N237 See Prior to Accident Subsection 15.3.3.3 Secondary Coolant Activity ANS-N237 See Table 15.0-9 Prior to Accident Total Steam Generator Tube 0.009 gpm 1 gpm l
Leak Rate During Accident and-Initial 8 Hours Activity Released to Reactor Coolant from Failed Fuel Noble Gas 0.0% of core 0.0% of core inventory inventory Iodine 0.0% of core 0.0% of core
-inventory inventory Iodine Partition Factor 0.0 0.01 l
Prior to the Accident Duration of Plant Cooldown 8
8 by Secondary System After Accident, hr.
Steam Release from 4 561,979 lb (0-2 hr) l Steam Generators 936,100 lb (2-8 hr)
Feedwater Flow to 4 793,091 lb 793,091 lb Steam Generators (0-2 hr)
(0-2 hr) 1,024,438 lb 1,024,438 lb (2-8 hr)
(2-8 hr)
Offsite Power Available Lost
- Per ANS-N237, American National Standard Source Term Specification.
- Condenser'available, steam released through condenser off-gas system at 60 SCFM.
15.3-18
~~
y
- ).
I.
TABLE 15.3-4
?.
ACTIVITY RELEASES TO' ATMOSPHERE FROM LOCKED ROTOR ACCIDENT
/
CONSERVATIVE ANALYSES' RELEASES (Ci)
REALISTIC ANALYSIS; PREACCIDENT ACCIDENT INITIATED
_ _. ~
ACTIVITY RELEASE (Ci)
. IODINE SPIKE IODINE SPIKE-Isotope-
'0-2 Er
~2-8 Hr 0-2 Hr 2-8 Hr 0-2 Hr 2-8 Hr:
I-131 6.0 (-4) 1.3 (-3) 2.5 (-1) 9.7 (-1) 2.0 (-1) l ~. 0 (+0)
I-132 8.9.(-S).
7.2. (-5) 6.6 (-1) 8.0 (-1) 9.1 (-1)
.4.3
(+ 0 )
I-133
- 7.1 ( 4) 1.3 (-3) 3.9 (-1) 1.3
(+'0 )
3.3 (-1)
'1.9
(+ 0 )
I-134
- 9. 2. (-60 2.9 (-6) 2.8 ( 2).
17.5 (-3) 6.1 (-2) 1.3 (-1)
I-135 2.2 (-4)
' 3. 6 (-4) 2.0 (-1):
5.1 (-1) 2.0 (-1) 1.2 (FO) wv Xe-133
'l.9 (-2) 5.7 (-2)'
l.71 (+1) 5.1 (+1) 7
-Xe-133m
- 3. 9 (-4) 1.1 (-3) 3.5 (-1) 9.9 (-1) l u,
4 Xe-135' l.1 (-3) 2.3 (-3) 9.9 (-1)-
2.1. (+ 0 )
"y.
l e
XE-135m 9.9 (-6)
. NEGLIGIBLE 8.9 (-3)
Xe and Kr Isotopes t
XE-138 3.6 (-5)
NEGLIGIBLE 3.2 (-2)
Same As PreAccident Spike Case Kr-85 2.7 (-5) 8.1 - (-5) '
2.4 (-2) 7.3 (-2)
Kr-85m 3.3 (-4)
- 5. 51 (-4 )
3.0 (-1) 5.0 (-1)
Kr-87 1.4 ( -4) 6.9 (-5) 1.3 (-1) 6.2 (-2)
Kr-88 5.9 (-4) 7.0 (-4) 5.3 (-1) 6.3 (-1)
~4 Note:
- 6. 8 (-4 ) = 6.8'x 10 h
a