ML20210U728
| ML20210U728 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 05/28/1986 |
| From: | Taylor J DAIRYLAND POWER COOPERATIVE |
| To: | Zwolinski J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM LAC-11629, TAC-44450, NUDOCS 8606030063 | |
| Download: ML20210U728 (9) | |
Text
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D DA/RYLAND h
[k COOPERAT/VE po Boxai7 2615 EAST AV SOUTH. LA CROSSE WISCONSIN 54601 (608) 788 4000 May 28, 1986 In reply, please refer to LAC-Il629 DOCKET NO. 50-409 Mr. John Zwolinski, Director BWR Project Directorate #1 Division of BWR ~ Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 POST-ACCIDENT SAMPLING SYSTEM (PASS) (TAC 44450)
NUREG-0737. ITEM II.B.3. CRITERION 2
References:
(1) NRC Letter, Zwolinski to Taylor, dated 4/10/86 (2) DPC Letter, Taylor to Zwolinski, LAC-11396, dated 2/4/86
Dear Mr. Zwolinski:
This submittal is in response to your letter of April 10, 1986 (Reference 1),
regarding development of procedures to improve methods to determine a more realistic estimate of core damage in post-accident situations in accordance with NUREG-0737 Item II.B.3, Criterion 2.
We have developed methodology to relate containment building high range area radiation monitor readings, at varying times after reactor shutdown, to potential fuel damage in a LOCA condition. This methodology is being incorporated into two procedures, EPP-6 and Operations Manual Volume X, Section 5.
Appropriate plant personnel will be trained on these changes by June 30, 1986. This methodology is found in Attachment 1 to this letter.
We have also developed methodology to relate reactor coolant fission product activities to potential fuel damage conditions using the General Electric NEDO-22215 information as guidance.
This methodology is found in.
This information is being incorporated into our procedure, EPP-6.
Training for appropriate plant personnel on this procedure change will be completed by June 30, 1986.
Containment building hydrogen analyzers are installed and would be used in a post-accident situation, as described in Operations Manual Volume XI, Section 6.5.2.
8606030063 860528 PDR ADOCK 05000409 p
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Mr. John A. Zwolinski, Director May 28, 1986 Division of BWR Licensing LAC-Il629 U.S. Nuclear Regulatory Commission page 2 Our post-accident sampling locations are described in the Operations Manual Volume XI, Section 6.'
We do not have local core temperature monitoring.
If you have any additional questions or coments, please contact us.
Very truly yours, DAIRYLAND POWER C PERATIVE James W. Taylor, General Manager JWT:PWS:dh Attachments cc - J. G. Keppler, Reg. Admin., NRC-DRO III NRC Resident Inspector J. Stang, Project Manager PC4-4 j
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ATTACHMENT A TABLE 2 HIGH RANGE CONTAINMENT BUILDING AREA RADIATION. MONITOR READINGS WITH 100% FUEL INVENTORY FISSION PRODUCT RELEASESlal INTO. CONTAINMENT AT VARIOUS TIMES AFTER REACTOR SHUTDOWNLDJ TIME AFTER APPROXIMATE REACTOR CONTAINMENT HRCBARM SHUTDOWN CONCENTRATIONS INDICATIONS (Hours)
(Ci/m3)
( R/hr) 0 6,050 1.33 E6 0.5 5,400 1.14 E6 0.75 5,200 1.08 E6 1.0 5,000 1.03 E6 2.0 4,500 8.70 ES 3.0 4,100 6.50 ES 6.0 3,400 2.90 ES 9.0 3,000 1.40 E5 12.0 2,700 9.0 E4 15.0 2,500 7.0 E4 18.0 2,300 6.0 E4 24.0 2,100 5.0 E4 36.0 1,800 4.0 E4 40.0 1,600 3.5 E4 96.0 1,300 3.0 E4 120.0 950 2.0 E4 240.0 460 1.0 E4 360.0 250 5.0 E3 480.0 150 3.0 E3 600.0 100 2.0 E3 720.0 80 1.7 E3 960.0 50 1.1 E3 1,200.0 43 9.2 E2 1,440.0 42 9.0 E2 (a)
Assonou that 100 percent of fuel inventory noble gases, 25 percent of fuel inventory of radiciodines and 1 percent of fuelinventoryofparticulatestranslocagefromfuelto 3
containment free air spaces (7.48 x 10 m ) and mix uniformly with containment isolated and functioning with design basis leak rate of 0.01 percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(b)
Assumes the longest exposure time occurs for the equilibirum core, which has an approximate exposure of 15,000 Mwd /ST for the central third of fuel, 10,000 Mwd /ST for the intermediate third, and 5,000 Mwd /ST for the outer third.
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NON CONTROLLED COPY DRAFT-REFERENG ONI.V ATTACHMENT A TABLE 4 ESTIMATION OF FUEL FISSION PRODUCT RELEASES (l)
BASED UPON HRCBARM READINGS AT DIFFERENTIAL TIMES AFTER REACTOR SHUTDOWN HIGH RANGE CB i
AREA RADIATION MONITOR READINGS lE6 R/HR SES R/HR lES R/HR SE4 R/HR lE4 R/HR SE3 R/HR lE3 R/HR SE2 R/HR 1
30 Minutes 90%
404 7.5%
4.0%
0.75%
0.4%
0.09%
0.04%
1 Hour 1004 50t 10.0%
5.0%
1.00%
0.5%
0.10%
0.054.
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60%
15.0%
7.5%
1.00%
0.5%
0.10%
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l 3 Hours 75%
20.0%
10.0%
2.00%
1.0%
0.20%
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0.30%
0.09%
9 70.0%
35.0%
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0.70%
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55.0%
10.004 5.5%
1.00%
0.70%
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15.004 7.0%
1.40%
0.80%
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1 Day 100.0%
20.00%
10.0%
2.004 1.304 2 Days 30.00%
15.0%
2.504 1.40%
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l 10 Days 100.004 50.0%
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20.00%
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30 Days 100.004 30.00%
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40 Days 45.00%
1 50 Days 55.00t s,
I 100 Days 58.00%
1 j
(1) 1004 FUEL FISSION PRODUCT RELEASES E 100% NOBLE ~GAES, 25% RADIOIODINES, 1% PARTICULATES.
l (2)
FUEL GAP ACTIVITY = THE ACTIVITY FOUND BETWEEN'THE FUEL PELLET AND INTERNAL SURFACES OF CLADDING AND ACTIVITY FOUND IN GAP SPACES BETWEEN FUEL PELLETS.
APPROXIMATELY 0.075% OF NOBLE
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GASES, 0.025% OF RADIOIODINES AND 0.003% OF OTHER FISSION PRODUCTS.
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Qs; s,%%A ATTACHMENT C qg q q ( f-p f s TABLE 1 ESTIMATION OF POST-ACCIDENT FUEL CONDITIONS x.. ' L91//
BY REACTOR COOLANT ANALYSIS FISSION PRODUCT CHEMICAL CLASS RATIO METHOD KEY RADIONUCLIDE RATIO RANGES AND NOMINAL AVERAGES (In Parentheses)
TYPE OF FUEL Cs-137 Cs-137 Te-132 Cs-137 Rb-88 Cs-134 I-131 Tc-99m CONDITION RELEASE Ru-103 Ce-141 Ru-103 Np-239 Sr-91 Zr-95 Ru-103 Sb-127 FUEL / CLADDING 10-50 110-500 2-5 3500-15500 3-10 40-170 8-30 3-7 MELTING RELEASE (35)
(325)
(5)
(9800)
(5)
(110)
(18)
(3) 1 OXIDATION No Data No Data 0.4-0.7 No Data No Data No Data 0.06-0.1 20-40 RELEASE (0.6)
(0.07)
(25)
VAPORIZATION 10-250 5-120 10-300 140-3700 3-56 No Data 0.5-10 0.005-0.1 RELEASE (52)
(25)
(60)
(700)
(13)
(1)
(0.02)
FUEL GAP ACTIVITY No Data No Data No Data No Data 370-No Data No Data No Data RELEASE 700,000 (24,000)
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WP33
Page 32 hkff DRidT -D...,,t. :
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t==ue 6 h~_L MON CONTROLLED COPY ATTACHMENT C TABLE 2 ESTIMATED COOLANT ACTIVITY CONCENTRATION RANGES (a)
FOR DIFFERENT FUEL PHYSICAL CONDITIONS (pCi/ml)
FUEL / CLAD FUEL GAP KEY INDICATOR MELT OXIDATION VAPORIZATION AC'fIVITY RADIONUCLIDE RELEASE RELEASE RELEASE RELEASE Rb-88 5.9E4-1.4E5 No Data 21.8E4 4.86E2-3.65E4
'Sr-91 5.283-5.2E4 No Data 5.0E2-1.lE4 7.0E4-9.9El Zr-95 3.9E2-3.983 No Data No Data No Data Tc-99m 3.9E3-3.9E4 2.9ES-3.6ES 3.7E2-8.9E3 No Data Ru-103 3.883-3.8E4 2.8ES-3.lE5 3.6E2-8.7E3 No Data Sb-127 1.2E3-5.693 8.lE3-1.6E4 8.lE4 6.9E3-9.6E2 I-131 1.lE5-2.2E5 1.7E4-2.lE4 2.2E3 2.2E2-4.4E4 Te-132 1.7E4-8.5E4 1.5E5-2.2E5 1.lES 9.4E2-1.3E4 Cs-134 6.6E4-1.5ES No Data 3.lE4 5.4E2-4.lE4 Cs-137 2.0ES-4.5E5 No Data 9.3E4 1.6E3-1.2E5 Ce-141 4.0E2-4.0E3 No Data 7.7E2-1.9E4 No Data Np-239 1.3El-1.3E2 No Data 2.5El-6.4E2 No Data I
(a)
Assumes activity is dispersed into 1.9E7 g (20.92 tons) of
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coolant water.
WP33