ML20198R470
| ML20198R470 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 11/03/1997 |
| From: | Hosmer J COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9711130287 | |
| Download: ML20198R470 (23) | |
Text
-.
Ormmonweilth Eden Compmy 14G) Opus Place.
Downen Grove, HJc515 5701 1
November 3,1997-U. S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D. C. 20555
Subject:
Response to Request for Additionr' k. formation Regarding the Revised Steam Generator Tube 4w re Analysis Byron Nuclear Power Station Facility Operating License NPF-37 and NPF-66 NRC Docket Numbers 50-454 and 50-455 Braidwood Nuclear Power Station Facility Operating License NPF-72 and NPF-77 NRC Docket Numbers: 50-456 and 50-457 y
References:
1.
T. Schuster Letter to USNRC, " Steam Generator Tube Rupture Analysis", dated April 25,1990 2.
USNRC Letter to Comed, " Safety Evaluation for Byron Units 1 and 2 and Braidwood Units 1 and 2 for the Steam Generator Tube Anslysis, dated April 23,1992 3.
J. Hosmer Letter to USNRC, " Steam Generator Tube Rupture Analysis for Byron and BraHwood Generating Stations", dated November 13,1996 4.
J. Hosmer Letter to USNRC," Steam Generator Tube Surveillance and Reactor Coolant System", dated February 28,1997 5.
J. Hosmer Letter to USNRC," Response to Request for Additional Information Regarding the Revised Steam Generator Tube Rupture A alysis'- Byron and Braidwood Stations", dated March 20,1997 6.
J. Hosmer Letter to USNRC," Response to Request for Additional Information Regarding the Revised Steam Generator Tube Rupture Analysis _- Byron and Braidwood Stations", dated June 24,1997 A.._ iOUl i
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USNRC November 3,1997 7.
J. Hosmer Letter to USNRC, " Response _to Request for Additional Information Regarding the Revised Steam Generator Tube Rupture Analysis - Byron and Braidwood Stations", dated August 19,1997.
8.
USNRC Letter to Comed, " Request for Additional Information Regarding the Revised Steam Generator Tube Rupture Analysis; Removal of Steam Generator Repair Methodologies; and Restoration of Previous Dose Equivalent lodine Limits - Byron and Braidwood Stations, dated October 3,1997 Via Reference 1, the Commonwealth Edison Company (Comed) transmitted the original Steam Generator Tube Rupture Analysis for Byron and Braidwood Units 1 and -
2 to the Nuclear Regulatory Commission (NRC). This submittal was approved in the Safety Evaluation transmitted via Reference 2. On November 13,1996, Comed subrnitied its revised Steam Generator Tube Rupture Analysis for Byron and Braidwood Stations due to the replacement of the Unit 1 steam generators (Reference 3) on February 28,1997. Comed submitted a request for a license amendment to revise the Technical Specifications regarding steam generator repair methodology and primary coolant dose equivalent lodine level (Reference 4). Comed provided responses to requests for additionalinformation (RAl) on March 20, June 24, and August 19,1997 (References 5,6 and 7). These RAls were in regards to the Reference 3 submittal.
Another request for additionalinformation was transmitted on October 3,1997 (Reference 8). This RAI addresses the submittals from both References 1 and 2. The attached document is Comed's response to the October,1997 requ.:st.
In the Reference 1 submittal, an auxiliary feedwater (AFW) flow rate of 464 gpm was assumed for the margin to overfill case. Comed has recently identified this assumption to be nonconservative. An operabiltiy evaluation has been performed for the current steam generators. It concluded that the Byron /Braidwood plants are stillin compliance with the licensing requirements established by the NRC for the mitigation of a Steam Generator Tube Rupture Event. For Byron Unit 1 and Braidwood Unit 1 modifications will be performed to limit the flow to 464 gpm as part of the replacement outages, B1R08 and A1R07, respectively. Plans for Byron Unit 2 and Braidwood Unit 2 will be dispositioned in B2R08 and A2R07, respectively.
' In addition, please note a typographical error in the November 13,1996, submittal (Reference 3). In Table 8 and Table 9 of the document, Reference 1 should be changed to Reference S. All other information in the tables remains the same.
(k:nla\\sgtriai4.dd2)
J
I
'USNRC November 3,1997 Plere direct any questions to this office, rh O / mnm John B. Hosmer Vice President Engineering JBHIML/rp Attachment cc:
Regional Administrator, Rill Byron /Braidwood Project Manager, NRR Senior Resident inspector, Byron Senior Resident inspector, Braidwood Office of Nuclear Safety, IDNS I
(k:nla\\sgtriai4. doc \\3)
Ll,
RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE REVISED STEAM GENERATOR TUBE RUPTURE ANALYSIS AND RESTORTION OF PREVIOUS DOSE EQUlVALENT IODINE LIMIT BYRON STATION AND BRAIDWOOD STATION (DATED 10/3/97)
The Byron and Braidwood submittals of the revised steam generator tube rupture (SGTR) analysis and the amendment request to increase the reactor coolant system (RCS) activity of dose equivalent *l are under review. Both submittals were made to support the replacement of the steam generators (SG) at Byron, Unit 1, and Braidwood, Unit 1. As a result of this review, it has been determined that there exists an insufficient amount of information to permit the staff to perform confirmatory calculations of the proposed actions. In order to complete these actions, the licensee is requested to provide information sufficient for the staff to model the conveyance and the release of radioactivity for the SGTR, main steamline break (MSLB), rod ejection and locked rotor accidents. If the replacement SG have no impact at all on the releases of radioactivity to the environment for any of these accidents, then data need not be provided for the accident involved. The licensee should supply the following information and any additional information that is necessary for the staff to accurately model the response of the replacement SG. For each of the accidents provide a time line for those aspects of the event relevant to the determination of releases to the environment.
- 1. For the MSLB accident, provide the following information:
- a. Mass of liquid released from the faulted SG as a function of time.
L For the purpose of dose calculation, the entire inventory of the faulted SG is assumed to boil dry in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
0 - 2 hrs 1.18E5 lb
- b. Mass of steam released from the intact SG as a function of time. As a minimum, releases should be designated as those within two hours and those after two hours.
(k:nlaisgtrrai4. doc \\4)
~.
. ~..
Total mass of steam released from the three intact SGs:
0 - 2 hrs 4.10E5 lb 2 - 8 hrs 9.49ES lb 8 - 16 hrs 6.77E5 lb 16-24 hrs 5.81E5 lb 24 - 32 hrs 5.22E5 lb 32 - 40 hrs 4.82E5 lb The steam masses indicated above for the replacement steam generators represent an increase over the original steam generator steam releases. The increase in steam release is required to cool the higher secondary side mass and to remove the higher primary liquid and metal sensible heat. Consistent with the Unit 2 methodology, these types of steam releases are divided over the period of time it takes to cooldown to RHR cut-in conditions,
- c. Flashing fraction for primary to secondary leakage into the intact SG.
0.0 (Unchanged from Unit 2 UFSAR analysis)
The intact SGs do not steam dry. An iodine partition factor of 0.1 is assumed in the intact SGs for the iodine carried over from the primary,
- d. Scrubbing fraction for flashed portion of primar); to secondary leakage into the intact SG.
0.0 (Unchanged from Unit 2 UFSAR analysis)
- e. Primary bypass fraction (liquid entrained in the flashing fraction) for intact SG.
0.0 (Unchanged from Unit 2 UFSAR analysis)
- f. Time to isolate faulted SG.
Two hours (Unchanged from Unit 2 UFSAR analysis)
The entire inventory of the faulted SG is assumed to boil dry in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (see item 1.a). All activity carried over to the fae!ted SG from the primary to secondary leakage is assumed to be released directly to the atmosphere during plant
- cooldown,
- g. Duration of plant cooldown by the secondary side.
40 hrs (Unchanged from Unit 2 UFSAR analysis)
(k:nla\\sgttrai4. doc \\5)
- h. Additional information which should be provided is contained in
. and Attachment 2.
See responses to Attachment 1 and Attachment 2 questions,
- 2. For the SGTR accident, provide the following information:
- a. Mass of liquid and steam released from the faulted SG as a function of time. As a minimum, releases should be designated as though within two hours and those after two hours.
There is no liquid release from the ruptured SG.
Steam Release from the faulted SG:
0-2 hrs 9.55E4 lb
> 2 hrs
- 0.0 lb
- The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate steam release during plant cooldown.
- b. Mass of steam released from the intact SG as a function of time. As a minimum, releases should be designated as though within two hours and those after two hours.
Total mars of steam released from the three intact SGs; 0-2 hrs 1.61ES lb
> 2 hrs
- 0.0 lb
- The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate steam release during plant cooldown.
- c. Flashing fraction in the intact and faulted SG.
See Table 1 for the ruptured SG flashing fraction. The primary coolant that flows through the ruptured tube flashes to steam and is conservatively assumed to carry the respective fraction of break flow nuclide activity concentration into the SG steam space.
For the intact SG, it is assumed that the primary coolant leakage is completely mixed in the secondary liquid and no flashing occurs.
(k:nla\\sgttrai4. doc \\6)
1
- d. Scrubbing fraction in the intact and faulted SG.
t 0.0 I
- e. Primary bypass fraction for the intact and faulted SG.
0.0
- f. Time to isolate faulted SG.
/
1881 seconds.
)
i lt should be noted that, the B/B SGTR methodology approved by the NRC (Ref.
- 1) does not calculate release during plant cooldown, including plant cooldown l
l that occurs while the ruptured SG flow is being isolated.
- g. Duration of plant cooldown by the secondary side.
The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate release during plant cooldown to RHR cut in. Therefore, this time is not provided.
- h. Primary to secondary release rate from the ruptured tube as a function of i
time.
See Table 2 for ruptured tube flow.
I.
Indicate if overfill conditions do or dr. not exist. If they do exist, appropriate mass release data should be provided as a function of time for the faulted SG.
The ruptured SG does not overfill.
J. Additionalinformation which should be provided is contained in.
See responses to Attachments 3 and 4.
l (k:nla\\sgttrai4. doc \\7) l
For the locked rotor accident, provide the following information:
3.
- a. Liquid release from the SG as a function of time.
Steam Releases (Total from four SGs):
0 - 2 hrs 5.65ES lb 2 - 8 hrs 1.089E6 lb 8 - 16 hrs 8.03E5 lb 16 - 24 hrs 7.07ES lb 24 - 32 hrs 6.48E5 lb 32 - 40 hrs 6.08E5 lb The steam masses indicated above for the replacement steam generators represent an increase over the original steam generator steam releases. The increase in steam release is required to cool the higher secondary side mass and to remove the higher primary liquid and metal sensible heat. Consistent with the Unit 2 methodology, these types of steam releases are divided over the period of time it takes to cooldown to RHf :ut-in conditions.
- b. Duration of plant cooldown by the seco%aiy side.
40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> (Unchanged from Unit 2 analysis)
- c. A description of how the primary to secondary releases were modeled as releases to the environment.
In the analysis, radionuclides carried by the primary coolant to the steam einerators, via the leaking tubes, are released to the environment via the steamline safety or power operated relief valves. (Unchanged from Unit 2 ana:ysis.)
The leak rate assumed in the Unit 2 analysis was 1 gpm (1440 gpd) while the maximum leak rate allowed by Technical Specification 3.4.6.2.c is 600 gpd total through all SGs (Ref. 2).
- d. Fraction of fuel rods experiencing cladding perforation and/or fuel melting.
i 5% (Unchanged from Unit 2 analysis)
I
'i 1
j (k:nla\\sgttrai4. doc \\8)
l 4.
For the rod ejection accident, provide the following information:
- a. The fraction of the fuel rods which have their cladding breached as a result of this accident.
10% (Unchanged from Unit 2 analysis)
- b. The fraction of the fuel rods which reach or exceed the initiation temperature for fuel melting as a result of this accident.
5% (Unchangt.d from Unit 2 analysis)
- c. For the release via the primary to secondary leakage pathway, a description of the assumptions which were utilized in the release of such activity.
Unchanged from Unit 2 analysis, except for the amount of steam release through the relief valves. For the RSG, the amount of release is calculated to be 116346 lb (total from four SGs).
The leak rate assumed in the Unit 2 analysis was 1 gpm (1440 gpd) while the maximum leak rate allowed by Technical Specification 3.4.6.2.c is 600 gpd total through all SGs (Ref. 2),
- d. For the release via the containment pathway, a description of the assumptions which were utilized in the release of such activity.
Unchanged from Unit 2 analysis (see B/B UFSAR, p.15.4-43 and 15.4-44).
(k:nla\\sgttrai4. doc \\9)
l Table 1 - F! ashing Fraction for the Ruptured SG Time (sec)
Flashing Fraction (Ib flashed /lb break flow) 0.00E+00 0
4.50E+01 4.67E-02 1.05E+02 4.65E-02 1.65E+02 4.65E-02 2.05E+02 4.67E-02 2.45E+02 4.69E-02 3.05E+02 4.74E-02 3.45E+02 4.75E-02 4.05E+02 4.74E-02 4.45E+02 4.76E-02 5.05E+02 4.75E-02 5.45E+02 4.77E 02 6.05E+02 4 81E-02
~
6.45E+02 4.80E-02 7.10E+02 1.13E-02 7.50E+02 1.88E-02 8.10E+02 2.67E-02 8.50E+02 2.91E 02 9.10E+02 3.12E-02 9.50E+02 3.20E 02 1.01 E+03 3.27E-02 1.11 E+03 3.28E-02 1.21 E+03 3.23E-02 1.31E+03 3.17E-02 1.41E+03 3.10E-02 1.51E+03 3.03E-02 1.61E+03 2.95E-02 1.71 E+03 2.86E-02 1.81 E+03 2.79E-02 1.91E+03 2.51 E-02 2.01E+03 1.68E-02 2.11 E+03
.1.17E-02 2.21 E+03 1.02E-02
- 2.31E+03 9.30E-03 2.41 E+03 8.49E-03 2.51 E+03 7.68E-03
~
2.61E+03 710E-03 2.71E+03 6.95E-03 2.81E+03 6.86E-03
~
2.90E+03 6.72E-03 3.00E+03 6.44E-03 3.10E+03 5.90E-03 3.20E+03 1.52E-03 I
3.30E+03 0.00E+00 (k:nla\\sgttrai4. doc \\l0)
Table 2 - Ruptured Tube Flow Time (sec)
Break Flow (Ib/sec) 0.00E+00 -
0 4.50E+01 4.62E+01 1.05E+02 4.55E+01 1.65E+02 4.47E+01 2.05E+02 4.42E+01
~
2.45E+02 4.37E+01 3.05E+02 4.31 E+01 3.45E+02 4.26E+01 4.05E+02 4.19E+01 4.45E+02 4.14E+01 5.05E+02 4.06E+01 5.45E+02 4.01 E+01 6.05E+02 3.94E+01 6.45E+02 3.89E+01 7.10E+02 3.64E+01 7.50E+02 3.91 E+01 8.10E+02 4.12E+01 8.50E+02 4.21 E+01 9.10E+02 4.32E+01 9.50E+02 4.37E+01 1.01 E+03 4.41E+01 1.11 E+03 4.45E+01 1.21 E+03 4.46E+ 01 1.31 E+03 4.48E+01 1.41 E+03 4.50E+01 1.51 E+03 4.52E+01 1.61E+03 4.55E+01 1.71 E+03 4.58E+01 1.81 E+03 4.62E+01 1.91 E+03 4.60E+01 2.01 E+03 4.48E+01 2.11E+03 4.51 E+01 2.21E+03 4.62E+01 2.31 E+03 4.71 E+01 2.41 E+03 4.78E+01 2.51 E+03 4.83E+01 2.61E+03 -
4.87E 501
. 2.71E+03 4.87E+01 2.81 E+03 4.85E+01 2.90E+03 4.81E+01 i
3.00E+03 4.74E+01
~
3.10E+03 4.38E+01 3.20E+03 3.98E+01 3.30E+03 3.74E+01
- (k:nla\\sgtriai4. doc \\l1)
References:
1.
" Byron, Units 1 and 2, and Braidwood, Units 1 and 2 - Steam generator Tube Rupture Analysis _ (Tac Nos. M57080, M63247,' M64026 and M64053)," letter from R. M. Pulsifer to T.J. Kovach, April 23,1992.
2.
Byron /Braidwood Technical Specifications, Bases Section 3/4.4.6.2.
(k:nla\\sgttrai4. doc \\l2) i
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i ATTACHMENT 1 INPUT PARAMETERS FOR EVALUATION OF MAIN STEAMLINE BREAK ACCIDENT
- 1. Primary coolant concentration for technical specification's (TS) maximum instantaneous value for dose equivalent *l.
Unchanged from Unit 2 analysis Pre-existina Solke Value (uCila)
- l = 38.7 132 = 43.4 1
- l = 61.9
- l = 9.3
- l = 34.0
- 2. Volume of primary coolant and secondary coolant.
Primary Coolant Volume (ft')
12062*
Primary Coolant Temperature ( F) 586.2 Secondary Coolant Steam Volume (ft )
2780 Secondary Coolant Liquid Volume (ft )
2423 Secondary Coolant Steam Temperature (*F) 549.2 Secondary Coolant Feedwater Temperature ( F) 440**
The same value as Unit 2 is used. There is no impact on the pre-existing spike case. Using the lower Unit 2 primary coolant volume in the Unit 1 accident initiated spike case is conservative because the release rate is higher with a lower primary coolant volume.
Main feedwater temperature.
3.
TS limits for DE *l in the primary and secondary coolant.
Primcry Coolant DE *l concentration
- 1. 0 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, TS value (pCi/g)
Secondary Coolant DE *l concentration ( Ci/g) 0.1 (k:nla\\sgttral4. doc \\l3)
4.
TS value for the primary to secondary leak rate.
Primary to secondary leak rate, any SG (gpd) 150 Primary to secondary leak rate, total all SGs (gpd) 600 The leak rate assumed in the Unit 2 analysis was 1 gpm (1440 gpd) while the maximum leak rate allowed by Technical Specification 3.4.6.2.c is 600 gpd total through all SGs (Ref. 2).
Primary to Secondary Leakage Faulted SG 1.0 Intact SG 0.1
- 6. Steam Released to the environment Faulted SG:
l For the purpose of dose calculation, the entire inventory of the faulted SG is y
assumed to boil dry in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
0 - 2 hrs 1.18E5 lb Total releases from the three intact SG:
0 - 2 hrs 4.10E5 lb 2 - 8 hrs 9.49E5 lb 8 - 16 hrs 6.77E5 lb 16 - 24 hr 5.81E5 lb 24 - 32 hrs 5.22E5 lb 32 - 40 hrs 4.82E5 lb
- 7. Letdown Flow Rate (gpm) 75 gpm (Unchanged from Unit 2 analysis)
This parameter is used in the iodine spike model only.
(k:nla\\sgttrai4. doc \\l4)
j 8.- Release Rate for 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> TS of Dose Equivalent '8'l i
Unchanged from Unit 2 analysis (includes a factor of 500):
GAlf
'8'l = 5.2E3 132 = 3.2E4 1
'88 = 1.2E4 1
- l = 1.6E4
5 = 1.2E4 1
- 9. Atmospheric Dispersion Factors (socim')
Unchanged from Unit 2 analysis.
Byron Braidwood EAB (0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 5.7E-4 7.7E-4 LPZ (0-8 hours) 1.7E-5 7.1 E-5 Control Room (0-8 hours) 4.05E-3 6.24E-3
- 10. Control Room Byrorl Braidwood Volume (ft')
4.05E5 4.05E5 Normal Makeup Flow (cfm)
Emergency Makeup Floa (cfm) 3.0E3 6.0E3 Makeup Filter efficienc) (%)
99 99 Unfiltered inleakage (cfm) 78.75 15.0 Recirculation Filter Flov t Rate (cfm) 4.5E4 4.5E4 Recirculation Filter Efficency (%)
90 90 Normal makeup is isolated during MSLB.
- i (k
- nla\\sgttrai4. doc \\l5)
ATTACHMENT 2 MAIN STEAMLINE BREAK THYROID DOSE ASSESSMENT BYRON UNIT 1 Pre-existina Spike EAB LP4 Control Room Calculated doses (rem) 2.77" 0.295" Regulatory Guidelines 300 300 30 (rem)
Accident initiated Spike EAB LP_Z Control Room Calculated doses (rem) 2.22" 0.309" Regulatory Guidelines 30 30 30 (rem)
The LOCA control room dose case bounds the main steamline break control room case. (See Comed to NRC Letter,"Additionalinformation Pertaining to the Tec'inical Specification Amendment for the Reduction in Dose Equivalent lodlie," dated October 1,1997.)
An inconsistency in the Byron /Braidwood UFSAR regarding the activity source term for the primary and secondary coolant has recently been identified. The calculated doses presented need to be reevaluated. The changes are expected to be minor.
1 (k:nla\\sgttral4. doc \\l6) l
MAIN STEAMLINE BREAK THYROID DOSE ASSES? MENT-BRAIDWOOD UNIT 1 Pre-existina Soike-EAB jfZ Control Room Calculated doses (rem) 3.84 "
1.32 "
Regulatory Guidelines 300 300 30 (rem)
Accident Initiated Soike EAB 12Z Control Room Calculated doses (rem) 3.10" 1.39" Regulatory Guidelines 30 30 30 (rem)
The LOCA control room dose case bounds the main steamline break control room case. (See Comed to NRC Letter, " Additional Information Pertaining to the Technical Specification Amendment for the Reduction in Dose Equivalent lodine," dated October 1,1997 )
An inconsistency in the Byron /Braidwood UFSAR regarding the activity source term for the primary and secondary coolant has recently been identified. The calculated doses presented need to be reevaluated. The changes are expected to be minor, i
(k:nla\\sgttrai4. doc \\l7)
ATTACHMENT 3 IWPUT PARAMETERS FOR EVALUATION OF SGTR 1.
Primary coolant concentration of TS value for dose equivalent l.
Pre-existina Spike Value (uCi/a)
'8'l = 38.7 132 = 43.4 1
'83 = 61.9 1
'3dl= 9.3
'85 = 34.0 1
2.
Volume of primary coolant and secondary coolant.
3 Primary Coolant Volume (ft )
11386. excluding pressurizer Primary Coolant Temperature (*F) 567 3
Secondary Coolant Steam Volume Total (ft )
12076*
Secondary Coolant Mass Total (Ibs) 443180*
Primary Coolant Pressure (psia) 2293 Primary Coolant Mass (Ibs) 538361 Pressurizer Volume (it )
1150" Pressurizer Temperature (*F) 657 Pressurizer Pressure (psia) 2293 Secondary Coolant Liquid Mass /SG (Ibs) 105224*
Secondary Coolant Steam Mass /SG (Ibs) 5571*
Secondary Steam Temperature (*F) 523 Secondary Liquid Temperature (*F) 523 Based on 55% narrow range SG level Based on 65% pressurizer level 3.
TS limits for DE *l in the primary and secondary coolants:
Maximum instantaneous in primary coolant ( Cilg) 60.0 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> DE ' 'l in primary coolant (pCi/g) 1.0 Secondary Coolant ( Ci/g) 0.1 (k:nla\\sgtrial4. doc \\l8) 1
ATTACHMENT 4' STEAM GENERATOR TUBE RUPTURE THYROlD DOSE ASSESSMENT (BYRON UNIT 1)
Case involvina Pre-existina Soike EAB LPZ Control Room Calculated doses (rem) 14,71**
0.44" not analyzed' Regulatory Guidelines (rem) 300 300 30 Case involvina Accident Initiated Spike EAB LPZ Control Room Calculated doses (rem) 13,82" 0.41" not analyzed
- Regulatory Guidelines (rem) 30 30 30 The LOCA control room dose case bounds the steam generator tube rupture control room case.
An inconsistency in the Byron /Braidwood UFSAR regarding the activity source term for the primary and secondary coolant has recently been identified. The calculated doses presented need to be reevaluated. The changes are expected to be minor.
(k:nla\\sgttrai4.doci22)
i e
)
- +
- 4. TS value for the primary to secondary leak rate (include reference temperature and pressure).
I Any SG (gpd) 150.0-Total all SG (gpd) 600.0 Reference temperature (*F) 70,0 Reference pressure (psla) 14.7 5.
Primary coolant activity (Cl) due to a pre-existing spike:
'Si = 9.5E3 -
l 32 = 1.06E4 1
33l = 1.51E4 13dl = 2.3E3
'35 = 8.3E3 1
6.
Primary coolant activity levels (pCilg) for accident initiated spike.
131 = 2.5 1
$32 = 2.8 1
183 = 4.0 1
13dl = 0.6
'35 = 2.2 1
7.
Primary coolant concentration at maximum instantaneous value of 60 Cilg dose equivalent '3'i.
13'l = 38.7 132 = 43.4 1
'33 = G1.9 1
341= 9.3
'35 = 34.0 1
8.
Primary Coolant Activity (Cl) for Accident initiated Spike.
53'l = 6.11E2 132 = 6.84E2 1
'33 = 9.77E2 1
13dl = 1.47E2 135 = 5.37E2 1
l (t:nta\\sgtrrai4. doc \\t9)
- 9.
- lodine Partition Factor Faulted SG-0.01 Intact SG 0.01 Condenser not credited
- 10. Steam Released to the environment as a function of time:
Faulted SG (Ibs):
0-2 hours 9.55E4
> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
- 0.0 Total releases from three intact SG (Ibs):
0-2 hours 1.61ES
> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
- 0.0 The B/B SGTR methodology approved by the NRC (Ref.1) does not calculate release during plant cooldown.
- 11. Letdown Flow Rate (gpm) 0.0.
- 12. Atmospheric Dispersion Factors (socim3):
Byron praidwood EAB (0-2 hours) 5.7E-4 7.7E-4 LPZ (0-8 hours) 1.7E-5 7.1 E-5 Control Room (0-8 hours) 4.05E-3 6.24E-3
- 13. Control Room:
Byron Braidwood Emergency Makeup Flow (cfm) 6.0E3 6.0E3 Makeup Filter efficiency (%)
99 99 Unfiltered Inleakage (cfm) 78.75 15.0 Recirculation Filter Flow Rate (cfm) 4.5E4 4.5E4 Recirculation Filter Efficiency (%)
90 90 Occupancy Factor (0-1 day) 1.0 1.0 (k:nla\\sgttrai4. doc \\20)
- 14. > For the Accident Initiated Spike Case Release Rate (Ci/hr)
SOOX Release Rate (Ci/hr)*
- l = 41.18 2.06E4 132 = 62.71 3.14E4 1
'3'l = 92.16 4.61E4
- l = 108.0 5.40E4
'85 = 83.52 4.18E4 1
- The B/B SGTR methodology approved by the NRC (Ref.1) calculates the release rates based on escape coefficients. The resultant release rates are different than those used in the main steamline break dose calculations.
- 15. Flashing Fraction, Primary Bypass and Scrubbing Fraction as a function of time.
Flashing fraction for the ruptured SG is provided in Table 1. The primary coolant that flows through the ruptured tube flashes to steam and is conservatively assumed to carry the respective fraction of break flow nuclide activity concentration into the SG steam space.
For the intact SG, it is assumed that the primary coolant leakage is completely mixed in the secondary liquid and no flashing occurs.
Primary bypass is 0.0 throughout the transient.
Scrubbing fraction is 0.0 throughout the transient.
- 16. Mass release rate through the ruptured tube as a function of time.
See Table 2.
(k:nta\\sgttrai4. doc \\21)
o.
STEAM GENERATOR TUBE RUPTURE THYROID DOSE ASSESSMENT l
(BRAIDWOOD UNIT 1)
Case involvino Pre-existino Spike i
EAB LPl Control Room Calculated doses (rem) 19.87" 1.83" not analyzed
- Regulatory Guidelines (rem) 300 300 30 Case involvino Accident initiated Soike EAB LPZ Control Room Calculated doses (rem) 18.67**
1.72" not analyzed
- Regulatory Guidelines (rem) 30 30 30 The LOCA control room dose case bounds the steam generator tube rupture control room case.
Aa inconsistency in the Byron /Braidwood UFSAR regarding the activity source term for the primary and secondary coolant has recently been identified. The calculated doses presented need to be reevaluated. The changes are expected to be minor.
l (k:nla\\sgttral4. doc \\23)