ML20084M958
| ML20084M958 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 05/02/1984 |
| From: | Tramm T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 8548N, NUDOCS 8405160227 | |
| Download: ML20084M958 (8) | |
Text
' '.
CN Commonwealth Edison
/
) one First Nation:1 Pitzt. ChicJgo. litinois
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- ~ ' Address R; ply to. Post Othee Box 167 N
Chicago tilinois 60690 May 2, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Byron Generating Station Units 1 and 2 Braidwood Generating Station Units 1 and 2 i
Reactor Coolant Pump Transients NRC Docket Nos. 50-454, 50-455, 50-456 and 50-457 Reference (a): June 7,1982 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. Review of this information should close Confirma-l tory Issue 34 of the Byron SER.
In reference (a) we indicated that during a locked rotor tran-sient, steam releases associated with the a stuck open steam generator relief valve, safety valve or dump valve would not pose a radiological hazard.
In telephone discussions with the NRC Staff it was noted that the existing transient analysis predicts that a small percentage of the fuel rods will experience departure from nucleate boiling (DNB) during a locked rotor transient. Westinghouse predicted that none of the fuel rods would fail during such a transient.
The NRC is, however, not able at present to endorse the technial basis for that fuel failure analysis.
Any fuel rods which experience DNB should be presumed to fail.
In the locked rotor transient substantial offsite releases would be predicted l
for the case involving a stuck open secondary side valve, design basis steam generator leakage, and coincident loss of offsite power.
The locked rotor transient has now been reanalyzed for the Byron /
Braidwood units.
The new analysis shows that no fuel rods will experience DNB and no fuel rods will fail.
The offsite releases for this case would therefore be well within the 10 CFR 100 limits.
Appropriate changes to the FSAR will be made to document this analysis. Enclosed are advance copies of FSAR sections 15.3.3 and tables 15.3-3, 15.3-4, 15.0-9 and 15.0-10. Offsite doses tabulated in tables 15.0-11 and 15.012 are still being recalculated but they are expected to decrease. These changes will be incorporated into the FSAR at the earliest opportunity.
N ho E
N s
r-x 2-May 2, 1984'
. H. R. Denton Please direct further questions regarding this matter to this of fice.
One signed' original and fifteen' copies of this letter and the enclosures are provided for NRC review.
Very truly yours, Wt (E,(&:--
/
T. R. Tramm Nuclear Licensing Administrator im Enclost tes
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B/B-FSAR 15.3.3.3 Radilogical 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 suf ficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant.
As a result of the accident, radienuclides 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.
15.3.3.3.1 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 primary system which increases the iodine r'elease rate from the fuel to the primary coolant to a value 500 times greater than the release rate corresponding to the maximum equilibrium primary system iodine concentration of 1 pCi/ gram 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/ gram of Dose Equivalent I-131.
b.
The noble gas concentrxtions in the primary coolant are based on 1 percent defective fuel.
c.
The secondary coolant activity is based on the 0.E. of 0.1 pCi / gram 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 i
during the worst transient remains considerably less than 2700*F the
. core will remain in place and intact with no loss of core cooling capability.
5906Q: 10/041284 L
s 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 Poser Operated Relief Valve (PORVl_
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 effected 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.
I i
5906Q:10/041284
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B/B-FSAR B/B-FSAR TABLE 15.3-3 ASSUNPTIONS USED FOR THE LOCKED ROTOR ACCIDENT EXPECTED DESIGN Power 3565 3565 Fraction of Fuel with Defects 0.0012*
See Text, Section 15.3.3.3 Reactor Coolant Activity Prior to Accident ANS-N237 See Text, Section 15.3.3.3 Secondary Coolant Activity Prior to Accident ANS-N237 See Table 15.0-9 Total Steam Generator Tube Leak Rate Ouring Accident and Initial 8 Hours 0.009 gpm 1 gpm Activity Released to Reactor Coolant fran 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 Prior to the Accident 0.1 0.01 Duration of Plant Cooldown -
by Secondary System Af ter Accident, hr 8
8 Steam Release from 4 561,979 lb (0-2 hr)
' 936,100 lb (2-8 hr)
Feedwater Flow to 4, 793,091 (0-2 793,091 lb (0-2 hr)
Steam Generators br) 1,024,438 (2-8 1.024,438 lb (2-8 hr) hr)
.0ffsite Power Availab,le Lost Per ANS-237, American National Standard Source Term Specification.
Condenser available, steam released through condenser of f-gas system at 60 SCFM.
5906Q:lD/041384 L
B/B-FSAR TABLE 15.3-4 ACTIVITY RM E_ASES TO ATMOSPHERE FROM LOCKE0 ROTOR ACCIDENT CONSERVATIVE ANALYSES RELEASES (Ci)
REALISTIC ANALYSIS PRE-ACCIDENT ACCIDENT INITIATED ACTIVITY RELEASE (C1)
IODINE SPIKE IODINE SPIKE Isotope 0-2 Hr 2-8 Hr 0-2 Hr 2-8 Hr 0-2 Hr 2-8 Hr I -131 6.8 (-4) 1.3 ( -3) 2.5 (-1) 9.7 (-1) 2.0 (-1) 1.0 (+0)
!-132 8.9 (-5) 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 (10) 1-134 9.2 (-6) 2.9 (-6) 2.8 (-2) 7.5 (-3) 6.1 (-2) 1.3 (-1) 1-135 2.2 (-4) 3.6 (-4) 2.0 (-1) 5.1 (-1) 2.0 (-1) 1.2 (40)
Xe-133 1.9 (-2) 5.7 (-2) 1.71 (41) 5.,) (+1)
Xe-133m 3.9 (-4) 1.1 (-3) 3.5 (-1) 9.9 (-1)
Xe-135 1 1 (-3) 2.3 (-3) 9.9 (-1) 2.1 (+0)
Xe-135m 9.9 (-6)
NEGLIGIBLE 8.9 (-3)
Xe and Kr Isotopes Xe-138 3.6 (-5)
NEGLIGIBLE 3.2 (-2)
Same As Pre-Accident Sp_ike Case Kr-85 2.7 (-5) 8.1 (-5) 2.4 (-2) 7.3 (-2)
Kr-85m 3.3 (-4) 5.5 (-4) 3.0 (-1) 5.0 (-1)
Kr-87 1.4 (-4) 6.9 (-5) 1.3 (-1) 6.2 (-2)
Kr-88 5. ') (-4) 7.0 ( -4) 5.3 (-1) 6.3 (-1)
Note:
6.8(-4) = 6.8 x 10-4 4
5906Q: 10/041284
B/B-FSAR
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TABI2 15.0-9 SECONDARY COOLANT EQUILIBRIUM IODINE ACTIVITY SEW /DP/ACCR!fJPFAdS[MYS25.
on Gi/$$/ & 4
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Isotope
-v Concentration (uCi/gm) m de#A I-131
_A e-r-Ce 23f I-132
_ o. 4 7-3-7 0, / # 5 I-133 oveeos6 O<#/ 0 I-134
.o etts 4, #NI I-las
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B/B-FSAR AMENBMENT-30
. 7"CH 1341 TABLE 15.0-10
' REAC2OR COOL &NT EQUIL M UM IO AND NOB W
ACTIVIT2ES USED AN ACCIDENT DO4E JtNALYSES' ACTIVITY ISOTOPE (uCi/qm)
)
l I-131.
39 76 I-132 14.31 I-133 63.62.
I-134 9.54 I-135 34.99 Xe-133 398.03 Xe-133m 4.39 8.91 Xe-135
~ 0.283 Xe-135m.
Xe-138 0.992 Kr-85 12.46 Kr-85m.
2.97 Kr-87' E,....
l 70
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