ML20140G817
| ML20140G817 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 04/22/1997 |
| From: | Barron H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M90590, TAC-M90591, NUDOCS 9705120176 | |
| Download: ML20140G817 (5) | |
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. Dukelhuer Company H B. thnos McGuire NucIcar Genemtion ikpartment lice hesident WOOHapn Ferry Road (MGOllP)
(704)S 5 4800 Huntmedie NCBW9310 (M)S54809 Fax DUKEPOWER April 22, 1997
'U.
S. Nuclear. Regulatory Commission Attention: Document Control Desk Washington, DC 20555 4
Subject:
McGuire Nuclear Station, Units 1 and 2, Docket Nos. 50-369 and 370 Supplement to Replacement SG Proposed TS Amendment (TACs M90590 and M90591)
By letters dated September 30, 1994 and September 18, 1995, Duke Power Company submitted proposed revisions to the McGuire Units 1 and 2 Technical Specifications.
The proposed changes were required for replacement.of the steam generators for each unit.
On March 12,
- 1997, a
conference call was held between representatives of the NRC Staff-and Duke Power Company.
The NRC staff requested information describing changes in parameters that-affect ' calculated doses during a steam line break.
Estimated changes'in significant parameters for the : steam line break are provided in Attachment 1.
r On April 7,
- 1997, e.
cc.iference call was held between-
. representatives of the NRC Staff and Duke Power Company.
The Staff requested a summary of the changes to the' steam generator tube rupture dose analysis.
Draft revisions to' Updated - Final Safety Analysis Report Table 15-24 are provided as Attachment 2.
The '. Updated Final Safety Analysis Report description of steam line break will be revised as a part of the next update.
Please contact R. O. Sharpe at (704) 382-0956 if you have any l
qdestions.
O/
- Very truly yours, l
w H. B. Barron Attachments 9705
- = 120176 970422 T,
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L. A.
Reyes, Regional Administrator U.
S. Nuclear Regulatory Commission, Region II 101 Marietta Street, NW, Suite 2900 Atlanta, GA 30323 S. M.
Shaeffer Senior Resident Inspector McGuire Nuclear Station V. Nerses Project Manager, ONRR
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6 1
U.
S. Nuclear Regulatory Commission April 21, 1997 Page 3 bcc: G. W.
Hallman M.
R. Robinson R.
L.
Morgan G.
B.
Swindlehurst H.
E.
Vanpelt T.
R. Niggel M.
T.
Cash K.
L. Crane G. A.
Copp ELL (EC050)
File MC-1201.37-27
4 McGuire Nuclear Station Response to NRC Staff Questions Steam System Piping Failure Parameter Current Analysis Impact of RSG's Power Level (M4t) 3565 3479 (Note 1)
Percent of Fuel Defected 1
No Change Reactor Coolant Activity UFSAR Table 15-11 No Change S/G tube Leakrate (gpd) 500 135 (Note 2)
Range of Flashing
$1 No Change Fractions S/G Flow Restrictor (Ft")
1.374 1.374 S/G Secondary Volume (lbm)
(0% power) 116,000 144,500 (Note 3)
(100% power) 104,000 122,600 (Note 3)
Initial Steam Release from Defective S/G (lbm) (0-2 181,000 199,600 hr)
Steam Release from three nondefective S/G's(lbm)
(0-2 hr) 752,000 No Change (2-8 hr) 1,128,000 No Change Time Sequence of Events UFSAR Table 15-13 No Change EAB Thyroid Dose Results 7.2 Note 4 for the Pre-Existent Iodine Spike (REM)
EAB Thyroid Dose Results 5.3 Note 4 for the Concurrent Iodine Spike (REM) i Note 1: Not an RSG impact.
As McGuire UFSAR Ch 15 events are reanalyzed, the power level is assumed to be 102% instead of 104.5%.
102% is considered adequate to bound uncertainties in power level measurement.
Note 2: The S/G leakage reduction has been requested as a Technical Specification change.
]
Note 3: The Ch 15 thermal-hydraulic analysis is done at HZP; the 4
dose analysis is done at full power.
Note 4: With the requested reduction in S/G leakage, the resulting accident doses are judged to be significantly less than the current UFSAR Ch 15 results.
k'77/fC W/4[Al'i~ h McGuire Nuclear Station Appendix 15. Chapter 15 Tablea and Figures
. Table 15-24. Parameters for Steam Generator Tube Rupture Dose Analysis
- 1. Data and assumptions used to estimate radioactive source from postulated accidents
- a. Power level (MWt) 3 4 'l3 _ --3566- - -
- b. Percent of fuel defected gg I
- c. Steam generator tube leak rate prior to accidentjgpa+
/35 y
- d. Offsite power Not available 3
- e. Reactor coolant activity Table 15-11
- 2. Data and assumpdons used to estimate activity released i.
2
- a. Iodine partition factor during accident
.01 NI l
- b. Steam release from defective steam generator Obm)
~5tr;000-i
- c. Steam release from three nondefective steam generators Obm)
(0-2hr)
/
4t2:000-l (2-8 hr) 959/Do o 4 %000-j
- d. Reactor coolant released to the defective steam generator Obm)
M S,o/ 5 m ^~'
j-
- 3. Dispersion data
- a. Distance to exclusion area boundary (m) 762
- b. Distance to low population zone (m) 8850 3
- c. r/Q'at exclusion are a boundary (sec/m3) 9.0E-04 i
(02hr)
- d. xA at low population zone (sec/m3) 8.0E-05' i
(0-8hr).
)
- 4. Dose data
- a. Method of dose celculations Section 15.9
- b. Dose conversica assumptions Section 15.9 2
- c. Doses (Rem)
See Table 15-12 2
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