ML20211H264
| ML20211H264 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 10/01/1997 |
| From: | Hosmer J COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20211H271 | List: |
| References | |
| NUDOCS 9710060312 | |
| Download: ML20211H264 (3) | |
Text
o.mnmna ranh i amo n>ninen) 1400 Opt.k plat r Imw nor$ Grme.11 W1% Udi October 1,1997 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Additional Information Pertaining to the Technical Specification Amendment for the Reduction in Dose Equivalent lodine Ilyron and liraidwood Units I and 2 3
Hl(C DockcMimbers: 50:454. 455. 456 and 457
References:
1)
J. Ilosmer letter to the Nuclear Regulatory Commission dated August 21,1997, transmitting Additional Information for the Reduction in Dose Equivalent lodine 2)
J. Ilosmer letter to the Nuclear Regulatory Commission dated August 22,1997, transmitting Additional Information for the Reduction in Dose Equivalent lodine 3)
G. Stanley letter to the Nuclear Regulatory Commission dated September 2,1997, transmitting Technical Specification Amendment Request 4) hiecting between the Nuclear Regulatory Commission and the Commonwealth Edison Company dated September 4,1997 5)
Teleconference between the Nuclear Regulatory Commission and the Commonwealth Edison Company da'ed September 19,1997 References 1 and 2 transmitted the dose calculations for the exclusion area boundary, low population zone and the control room as the rest.lt of a hiain Steam Line lireak at Ilyron Unit 1. Reference 3 transtnitted the same calculations for Braidwood 1.
At the Referenced meeting, the Commonwealth Edison Company (Comed) and the Nuclear Regulatory Commission met to discuss these calculations. Subsequent to that
[
meeting, Comed has reviewed these calculations and has issued a revision to provide clarification along with correcting some infonnation. The attachment provides those revisions and specifically addresses:
For letdown and Reactor Coolant System (RCS) conditions, the appropriate density /
should be that ofcompressed water not saturated liquid. This was appropriately applied in the calculations transmitted in References 1,2 and 3. In the revised j
calculation,13yron changed the let down parameters from UFSAR Chapter 9 to
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Chapter 11 values.
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U.S. Document Control
-2 October 1,1997 UFSAR Tables ;5.14,15.1-4b are labeled incorrectly. Table 15.14 should be e
labeled," Activity Releases to Atmosphere from Steamline lin k Accident Unit 1."
applicable to litaldwood; and UFSAR Tablel5.1 4b should be labeled, " Activity Releases to Atmosphere From Steamline llreak Accident Unit 2," applicable to Ilyron and tiraidwood.
Referencing Tables 2.F.3 and 3.F.3, the Ilyron and llraidwood 0 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> weighted activity release was corrected.
A discrepancy between various UFSAR tables referencing primary and secondary concentrations for lodine 132 was resolved.
The control room dose calculation has also been revised to include the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> dose for both a Main Steam Line llreak (MStil) and Loss of Coolant Accident (LOCA).
4 Providing the dose for the same time period allows a more direct comparison of l
resultant doses.
Corrections to references and variable names.
These revisione did not change the allowable leak rate value or the conclusions reached in the original calculations submitted via References 1,2, and 3.
Additionally, the following addresses the questions raised by the Staff during the referenced teleconference.
Question 1: Does the flyron value stated on page 23, of the August 21' submittal for predicted leakage account for tube support plate as well as circumferential crack indications?
Response 1: Yes Question 2: At what time is it assumed that the faulted steam generator is isolated following an accident?
Response 2: Primary to secondary leakage in the faulted steam generator from a MSLil is considered to be isolated when the primary side is depressurized and there is no pressure differential across the steam generator tube wall. The primary side is initially depressurized by rapid cooling from the secondary side break flow and then through the use of the steam generator PORV's on the 3 intact loops until the Ril system can be started (375 psia) at approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Final depressurization and therefore isolation occurs at 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> into the event. The primary to secondary leakage is assumed to be constant at full difTerential pressure (2560 psid) for the entire event until isolation.
Question 3: Is the control room make-up flow rate the same for Ilyron and llraidwood?
Response 3: Yes, the outside air make-up flow rate is 6000 cfm k'nla bybwd stmgen ckmecal2 2
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U.S. Document Control 3-October 1,1997 Question 4: Regarding page 7 of Exhibit C of the Braidwood calculation (the September 2,1997, submittal), the Staff questioned that Reference 7 (UFSAR Table 6.4 1) does not provide the dose in the control room following a LOCA.
Hesponne 4: Comed's review of Reference 7 indicates that UFSAR Table 6.4 1,
" Expected Dose to Control Room Personnel at Braidwood Station Following a Loss of Coolant Accident (LOCA)" is appropriate. Additionally, the revised Braidwood calculation references the correct 50 day dose from Braldwood UFSAR Table 6A l.
Question 5:Was Reference 5 (BWR DIT 97 278) of the September 2,1997, submittal transmitted to the Staff?
Respamme 5: Reference 5 was not transmitted. BWR DIT 97 278 is a document that transmits information from one organization to another and contains information that is available elsewhere, if you have any questions, please contact this of0ce.
Sincerely,
,/ vo m John B. Ilosmer Engineering Vice President Attachments: A. Byron Off Site Dose Calculation B. Byron Control Room Dose Calculation C, Braidwood Off-Site Dose Calculation D. Braidwocxl Control Room Calculation ec:
A. Beach, Regional Administrator Rlil O. Dick, Byron Project Manager - NRR D. Lynch, Senior Project Manager NRR S. Burgess, Senior Resident inspector Byron C, Phillips, Senior Resident inspector Braidwood Of0cc of Nuclear Safety I
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