ML20003B611
| ML20003B611 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 02/02/1981 |
| From: | Ballentine J, Eddings M TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20003B610 | List: |
| References | |
| NUDOCS 8102250006 | |
| Download: ML20003B611 (47) | |
Text
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TENNESSEE VALLEY AUTHORITY DIVISION OF NUCLEAR P0hT.R SEQUOYAH NUCLEAR PLANT 4
1 MONTHLY OPERATIN6 REPORT JANUARY 1 - JANUARY 31, 1981 3
j DOCKET NUMBER 50-327 LICENSE NLHBER DPR-77 i
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Submitted By: ~
_s.3 fl 'Dh\b 8 Power E{aint Superintendent exon sse o dq,
o TABLE OF CONTENTS Operations Summary . . . . . . . .. . . . . . . . . . .. . . . .. . . .1 PORV's and Safety Valves Summary . . . . . . . . . . . . . . . . . . . . . 1 Offsite Dose Calculation Manual Changes . . . . . . . . . . . . . . . . . 2-11 Major Changes to Radioactive Waste Treatment System . . . .. .. .. . .12-15 Significant Operational Events . .. .. . . . . .. . . . .. . . . . . .16-17 Average Daily Unit Power Level . .. . . . . . . . . . . . . . . . . . . .18 Operating Data Report . .. . . . . . .. . ... . . . . . . . . . . . .19 Unit Shutdown and Power Reduction .. . . . . . . . .. . . . . . . . . .20-21 Plant Maintenance Summary . . . . . . . . . . . . . .. . .. .. . . . .22 Outage Maintenance Summary . . . . . . . . . . . . . . . . . . ... . . .23 Appendix A . .. . . . . .. . . . . . . . . . . . . . . . . . . .. . . .24-37 Appendix B . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .38-45 4
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Operations Summary January, 1981 The following summary describes the sigr ificant operational activities for the month of January. In support of this summary, a chronological log of significant events is included in this report.
There were sixteen Licensee Event Reports and no Special Reports durice the month.
There were seven scrams during January.
Unit I was critical for 496.06 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, produced 497,248 MWH (gross) with 5.81 percent station use, resulting in a average hourly gross load of 1,047,279 KW during.the month. The net heat rate for the month was 10,190 BTU /KWH. There are 3,454 full power days estimated remaining until the end of cycle 1 fuel.
With a capacity factor of 85 percent the target EOC exposure would be reached March 13, 1982. The capacity factor for the month was 70.5 percent.
At the end of the month, the reactor was in mode 1 at 96% power producing 1128 MW's.
PORV's and Safety Valves Summary No pressure operated relief or safety valves were challenged during the month of January.
Sequoyah Nuclear Plant Offsite Dose Calculation Manual Changes Changes in the Sequoyah Nuclear Plant ODCM are described in this section in accordance with Sequoyah Technical Specification 6.1.14.
Changes 1 through 10 were officially approved by the RARC on November 24, 1980. Change 11 is a typographical error that was noted by the RARC and approved for change September 26, 1980. Change 11 was withheld for reporting purpose until changes 1 through 10 were ready since it was only a typographical error and represented no real change in the ODCM methodology.
See Appendix A at the end of this report for the approved ODCM page changes.
Change 1 Description of Change Page 21 "The sum of the ratios at the diffuser pond....' should read "The sum of the ratios at the diffuser pipes. . . ."
Antlysis or Evaluation Justifying Change TVA has performed a study on the mixing effects of the Sequoyah Nuclear Plant diffuser pond and has determined that the diffuser pond can be used for dilution purposes. Effects of the diffuser pond on dilution are only to be considered when a unit is operating in the Cooling Tower " Helper" or "Open" mode.
Evaluation of Accuracy of Dose Calculations or Setpoint Determinations Accuracy of dose calculation or setpoint determination will not be affected by this change. This change only effects determination of concentra-tions of contamination released. Methodology for dose and setpoint determina-tion remains the same.
Change 2 Description of Change Page 21 - The following should be added at the end of 2.1.3: For releases into the cooling tower blowdown line, additional mixing is assumed to occur in the diffuser pond when SQN is operating in Helper or Open cooling mode.
Sequoyah Nuclear Plant Offsite Dose Calculation Manual Changes (Continued)
Description of Change (Continued)
The results, from field tests conducted in September 1979, are expressed in terms of relative concentration r:
r = 2.4x10~0 (F + fg+f2 #f 3*I) 4 Equation 2.3 then becomes F
ff l' )
- 2(R2 ~I)
- f3 ( 3'
- 2.4x10-6 (F + f7+f2+f3* 4) l_ i f 4(R4 -1)._
$F (2.3a)
Analysis on Evaluation Justifying Change See Change 1.
Evaluation of Accuracy of Dose Calculations or Setpoint Determination See Change 1.
Change 3 Description of Change Page 22a "...used in Equation 2.3..." should read "...used in Equation 2.3a..."
Analysis or Evaluation Justifying Change See Change 1.
Evaluation of Accurccy of Dose Calculations or Setpoint Determinations See Change 1.
a Sequoyah Nuclear Plant Offsite Dose Calculation Manual Changes (Continued)
Change 4 Description of Change Page 22a - Equation 2.5 should read:
2.4x10' (F + fg+f2*f3+ 4) #
f l(D11R -1) + f2 ( 2 2' f 3(b33R 'I) + I (b 4 44R -1) 5F Analysis or Evaluation Justifying Change See Change 1.
Evaluation of Accuracy of Dose Calculations or Setpoint Determinations See Change 1.
Change 5 Description of Change Page 23 "...for 2 release points and minimum dilution flow this be-comes..." should read "...for 2 release points, minimum dilution flow, and no diffuser pond dilution this becomes..."
l Analysis or Evaluation Justifying Change i.
See Change 1.
Evaluation of Accuracy of Dose Calculations or Setpoint Determinations See Change 1.
1 Sequoyah Nuclear Plant Offsite Dose Calculation Manual Changes (Continued)
Change 6 Description of Change Pages 24 and 25 - Equations 2.11, 2.12, 2.13, and associated changes should read:
5 1
D rem j k *.95 ,
(DCF)g) x I ik (2.11) i=1 where:
D jk = dose for the j organ from 5 radionuclides from the th k release point rem.
j = the organ of interest (thyroid or total body).
k = the release ptint of interest (cooling tower blowdown or turbine building sump).
.95 = conservative correctiot factor, considering only 5 radionuclides.
DCF.. = adult ingestion dose commitment facror for the j' organ 13.
h from the i radionuclide rem /pri, see attached as Table 2.1.
-Iik = m nthly activity ingested of the i th radionuclide from the th k release point, pCi.
Ig is described by-365 A Y ik I , pCi (2.12) ik = 12 o d
k a Sequoyac Nuclear Plant Offsite Dose Calculation Manual Changes (Continued)
Description of Change (Continued) where:
365 = days per year A ik = activity released of i' radionuclide during the month from th the k release point, pCi.
V = average rate of water consumption (730 ml/d ICRP 23, p. 358) 12 = months per year u = total monthly waste released and water for dilution, ml.
d = diffuser pipe dilution (5).
"ae dose equation then becomes 5
6 D (DCF)g xA ik mrem (2.13) jk = 4.67x10 u
i=1 Analysis or Evaluation Justifying Change
, This change expands the affected equations to account for multiple
[ release points. -No affect on the derivation or use of the equations results l from this change.
Evaluation of Accuracy of Dose Calculations or Setpoint Determinations l
l Accuracy of dose calculation or setpoint determination will not be affected by this change. This. change only expands the affected equations to allow for multiple release points. Methodology for dose and setpoint determination remains the same.
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. _ _ _ = . --
Sequoyah Nuclear Plant Offsite Dose Calculation Manual Changes (Continued)
Change 7 Description of Change Page 25 - The following should be added at the end of 2.3.2.1:
. For continuous releases into the cooling tower blowdowr. line, t
additional mixing occurs in the diffuser pond when SQN is operating in Helper or Open cooling mode. The relstive con-centration r, is given as a function of total flow into the pond:
r = 1.48x10'I' u Multiplying equation 2.13 by r, the dose from cooling tower blowdown becomes:
5 i
D k
- 6.906x10 -8 (DCF)g xA ik mrem (2.13b) i=1 Analysis or Evaluation Justifying Change See Change 1.
t.-
i Evaluation of Accuracy of Dose Calculations or Setpoint Determinations See Change 1.
l Change 8 Description of Change Pages 25 and 26 - Equations 2.14, 2.15, and associated changes should read:
I ik= Aik i , pCi (2.14) l ud
t Sequoyah Nuclear Plant Offsite Dose Calculation Manual Changes 1
(Continued)
Description of Change (Continued) where:
Aik = activity released of i radionuclide during the month from the k th release point, pCi th Ei E" B1 = fish concentration factor of i radionuclide pCi,ml, see attached as Table 2.1.
3 M = amount of fish eaten monthly (1.9x10 gm) u = total monthly waste released and water for dilution, al d = diffuser pipe dilution (5)
The dose equation then becomes 5
5 D.k J = 4x10 A.k i x B,i v LCF.1;mrem (2.15)
I i=1 l Analysis or Evaluation Justifying Change See Change 6.
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! Evaluation of Accuracy c f Dose Calculations or Setpoint Determinations See Change 6.
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Sequoyah Nuclear Plant Offsite Dose Calculation Manual Cnanges (Continued)
Change 9 Description of Change Page 26 - The following should be inserted af ter Equation 2.15:
For continuous releases into the cooling tower blowdown line, additional mixing occurs in the diffuser pond when SQN is operating in Helper or Open cooling mode. The relative con-centration r, is given as a function of total flow into the pond:
r = 1.48x10 -14 g Multiplying equation 2.15 by r, the dose from cooling tower blowdown becomes:
5
-9 jk = 5.912x10 A ik xB fx DCF g mrem (2.15b) i=1 Analysis or Evaluation Justifying Change See Change 1.
, Evaluation of Accu.. cy of Dose Calculations or Setpoint Determinations i
See Change 1.
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Change 10 Description of Change Pages 26 and 27 - Equations 2.16, 2.17, 2.18, 2.19, and associated
{ changes should read:
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Sequoyah Nuclear Plant Offsite Dose Calculatior. Manual Changes (Continued)
Description of Change (Continued) m D rem jk = ,
D ij k , (2.16) i=1 m
=
(DCF)1].. x I.k, i
rem (2.17) i=1 where:
t D ijk = dose to the j' organ from the i radionuclide from the k' release ,aint, rem.
J = the organ of inte est (bone, GI tract, thyroid, or total body).
th (DCF)13. . = adult ingestion dose commitment factor for the j l
organ from the i radionuclide, rem /pci, see Table 2.1.
th th Iik = activity ingested of the i radionuclide from the k release point, pCi.
I ik f r Water ingestion is described by:
A.
1 Vn pCi I. =
1 od (2.18) l-and for fish ingestion I. is described by i
pCi
! I = A.ki B.1 M ,
(2.19)
L i o d-4 Sequoyah Nuclear Plant Offsite Dose Calculation Manual Changes (Continued)
Description of Change (Continued) where th Aik = activity released of j radionuclide during the release th period from the k release point, pCi.
V = average rate of water consumption (730 ml/d).
n = number of days during the release period (d),
u = total mocthly waste released and water for dilution, ml.
d = credit for diffuser pipe dilution (5).
9 i/gm th Bf = fish concentration f actor of the i radionuclide, pCi/ml 3
M = amount of fish eaten monthly (1.9x10 gm)
Analysis or Evaluation Justifying Change See Change 6.
Evaluation of Accuracy of Dose Calculations or Setpoint Description See Change 6.
Change 11 Description of Change l 5 3 X/Q value on page 18 of the ODCM--The listed galue of 1.13x10 s/m is a misprint. The correct value'is 1.54x10 s/m . (This is applicable fo a real cow located in the north sector.)
Analysis or Evaluation' Justifying Change None required. Misprint.
Evaluation of Accuracy of Dose Calculations or Setpoint Descriptions l
No change made in dose calculation or setpoint description.
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Major Changes To Radioactive Waste Treatment System The major change in the Sequoyah Euclear Plant radioactive waste system described in this section is reported in accordance with Sequoyah Technical Specification 6.15.1, Major Changes to Radioactive Waste Treatment Systems.
The proposed change was reviewed by PORC on September 8, 1980, with approval recommended. Since the system installation has been delayed, official re-
- porting of the change has been postponed until the present. With installa-tion and testing scheduled for February,1981, the change is being reported on this January report.
Evaluation Summary The proposed installation of a portable demineralization system was reviewed in accordance with 10 CFR 50.59 using the Unreviewed Safety Question Determination (USQD) procedure. The USQD, approved by PORC on September 8, 1980, determined that the installation of the portable demineralization facility and associated plant modifications do not involve an unreviewed safety question.
Reason for System Change Sequoyah Nuclear Plant is equipped by design with Westinghouse evaporators for concentration of liquid wastes in ! s waste management system. The Westinghouse design has generated poor operating records at other nuclear plants. The evaporators installed at Sequoyah similarly have been shown to be unreliable and-ineffective during system testing.
In order to maintain plant availability, replacement equipment must be provided to handle those wastes which would normally (i.e., by design) be processed in the installed evaporators. A portable demineralization system has been selected due to its low cost, its relatively simple operation, and the minimal number of plant modifications / changes necessary to permit its installation and use.
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( System Description l
! The portable demineralization system is designed to remove both suspended and dissolved radioactive impurities contained in liquid wastes. The expected DF (decontamination factor) across the system is 100.
The proposed system contains two carbon-steel, disposable, epoxy-lined vessels (liners) having a vclume of approximately 150 cubic feet each, con-nected in series. The first vessel of the series contains a filtering medium whose primary purpose is to remove oil and grease from the waste stream, thereby protecting the resia from organic fouling. A positive-displacement, air-operated diaphragm pump takes suction on the vessel's effluent and discharges into the second vessel. This second vessel con-tains approximately 125 cubic feet of mixed-bed ion exchange resin which removes dissolved radionuclides from the waste stream. Waste water enters l
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Major Changes To Radioactive Waste Treatment System the vessel through an influent distribution header located at the top of the vessel. Waste water flows by gravity through the bed and exits the vessel through an effluent header located at the bottom of the vessel. A second diaphragm pump (identical to the one previously described) returns the processed water to the plant radwaste system for effluent discharge in a manner similar to that designed fc- the distillate from the waste evaporators.
The unit will be located in the truck access bay of the auxiliary building (controlled or regulated area) and will be monitered by an operator on a continuous basis when in operation. The system components will be housed within a concrete block retention dike lined with a vinyl material to preclude contamination spread due to potential leakage on vessel rupture.
The access bay contains floor drains to collect any liquid waste that may escape the system boundary due to dike failure. A small dike across the access door will confine such potential leakage to the regulated area. To limit personnel exposure, concrete culverts will be used as a shielding medium about each vessel liner.
Release and Exposure Evaluations Personnel Exposure Sequoyah Nuclear Plant Radiological Control Instructions and Health Physics Manual guidelines will be strictly followed during all operations of the demineralizer.
The radiation area created will be controlled and exposure kept to a minimum by use of adequate shielding (concrete culverts) to ensure that no high radiation area is created. The demineralizer will be operated in the truck access bay of the auxiliary building and no increase in exposure to persons in the unrestricted area will occur.
7 Exposure to plant personnel will actually be reduced by the use of l
portable demineralizers. The following data was developed, based on actual plant data from an operating Westinghouse 1100 MWe plant that switched from inhouse solidification of evaporator concentrates to portable demineralizers and reduced exposure in radwaste by approximately 30 man-rem /yr.
Sequoyah should reduce its exposure by 25.5 man-rem /yr based on worst case utilization of demineralizers.
Man-Rem Estimates Packaging and handling of evaporator concentrates Contact time - 5 minutes Average dose rate .2 R/HR 6732 concentrate -drums produced annually (Sequoyah FSAR) 5 min 6732 drums g g,
,60 min x .2 R/RR = 112.2 Man-rem /yr drum year l _
4 Major Changes To Radioactive Waste-Treatment System i
(Continued)
Personnel Exposure (Continued)
Operation and handling of portable demineralizers Contract time - 10 minutes Average dose rate - 10 R/HR Demineralizer used annually: (Max 52) (Expected 30) 10 m 52 demins 1 Man x x 10R{HR x = 86.7 Man-rem /yr 60m{nin demin year Minimum expected saving is 25.5 man-rro.
-112.2 man-rem
-86.7 man-rem 25.5 man-rem General Public Exposure No increase in radicactive liquid releases as defined in the FSAR or in exposure to the general public will occur by use of the portable demineralizer.
Hence, the releases from the plant via the demineralizer are in accordance with tbc design objectives of the waste processing system and Federal regulations.
4 4
Release Data
- The table shown below is a comparison of the predicted releases of liquid . [
radioactive materials, both with and without the use of the proposed portable demineralizer system to the actual releases for the months of October and November 1980.
8
- 1. _RADI0 ACTIVE LIQUID EFFLUENTS l
Predicted Concentration Predicted Concentration Actual Avg. Discharge 'from License using Portable
' Month Concentration, uc/m.1* Application, uc/ml** Demineralizer, uc/ml***
October 8.34 x-10 -07 6.62 x 10 -07 8.34 x 10 -09 1980 November. 176 x 10 -07 6.62-x 10 -07 1.76 x 10 -09 1980
! '*Value from monthly report.
- Based on value from Sequoyah FSAR, volume 9,_ table 11.2-7.
l ***Value from
- divided by the portable demineralizer decontamination factor
- of 100;
Major Changes To Radioactive Waste Treatment System (Continued)
, Volume of Radwaste Generated Design data 6,732 drums generated / year 7.5 cubic feet / drum j- 50,490 cubic feet / year waste generated Portable Demins (worst case) 52 demineralizers/ year 180 cubic feet /demineralizer 9,360 cubic feet / year waste generated 50,490 cubic feet / year -
Volume by evaporation and solidification 9,360 cubic feet / year -
Volume by demineralization 41,130 cubic feet / year -
Volume of waste not generated each year due to the use of demineralizers i
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Significant Operational Events Date Time Event 01/01/81 0001 Unit I was in mode 1 at 90% power, 1030 MW's.
01/03/81 2054 Reactor trip #30 occurred when power was restored to the generator transformer relay auxiliaries causing the EHC system to switch back to load control at the minimum load setting. This caused an approximate 81%
load rejection which resulted in the Lo-Lo steam generator level condition that tripped the reactor.
01/04/81 1118 Reactor critical.
01/09/81 2250 Began power reduction for maintenance on main feedwater pumps.
01/11/81 0555 Power reduction ended, began increasing power.
01/14/81 1921 Began S.U.-10.1 Test - 300 HR, 100% power test.
01/15/81 1001 Reactor trip #31 occurred when transfe:rring the A $ main transformer cooling supply to normal feed. This caused a low flow alarm on the low voltage bushing oil pump which energized the 180-82 relay. The 30 second delay on the 180-82 relay did not function properly resulting in a generator trip and subsequent reactor trip, i
01/15/81 2133 Reactor critical.
l 01/18/81 0418 Reactor trip #32 occurred when oscillations occurred in the EHC control system governor valves position limiters. The governor l valves went closed and the steam generator i
levels dropped resulting in the reactor trip.
01/18/81 0645 Reactor critical.
01/18/81 1307- Reactor trip #33 occurred due to a Lo-Lo steam generator level in steam generator
- 4 during recovery from reactor trip #32.
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Significant Operational Events Dat,e Time Event 01/18/81 1710 Reactor critical.
01/18/81 2002 Attempted to tie the generator online 160 degrees out of phase. Received a generator trip from the generator back-up relay.
01/19/81 0352 Reactor trip #34 occurred when a control rod urgent failure alarm was received in the control rod system. This alarm was received when oscillation started occurring in the EHC control system governor valve position limiter, other plant instrumentation and in the steam generator levels. The UO manually tripped the reactor.
01/20/81 0024 Entered Mode 4.
01/27/81 1620 Reactor critical.
01/27/81 1958 Reactor trip #35 occurred due to a Lo-Lo steam generator level in steam generator #3 caused by low air pressure affecting the operation of a feedwater level control valve.
01/28/81 0250 Reactor critical.
01/28/81 0557 Reactor trip #36 occurred when steam dump valve 1-FCV-1-111 failed open causing a Hi-High level on steam generator #2 causing a turbine trip and subsequent reactor trip.
01/28/81 0947 Reactor critical.
01/31/81 2400 Unit I was in Mode 1 at 96% power; 1128 MW's.
AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-327 UNIT 1 DATE 2-2-81 COMPLETED BY Michael Eddings TELEPHONE (615) 842-0295 s
MONTH January 1981 DAY -AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)- (MWe-Net) 1 '1000 17 671 2' 909 18 485 3- 899 - 19 218 4 '573 20- 0 5 907 21 0 6 1009 22 0 7 1991 23 0 8 1048 '24 0
.9 1059 25 0 10' 626 26- 3 11 956 27 0 12 1156 28 282 13 1097 29 1011 14 1123 30 1083 15 429 31 1085 16 _
1049 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.
(9/77)
OIERATING DATA REPORT DOCKET NO. so_177 DATE 2-2-81 COMPLETED Bi Michael G. Eddings TELEPHONE (615) 842-0295 OPERATING STATUS
- 1. Unit Name: Sequovah one Notes
- 2. Reporting Period: Januarv. 1981
- 3. Licensed Thermal Power (hWt): 3411
- 4. Nameplate Rating (Gross MWe): 1220.58
- 5. Design Electrical Rating (Net MWe): 1148
- 6. Maximum Dependable Capacity (Gross MWe): 1183
- 7. Maximum Dependable Capacity (New MWe): 1148
- 8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report, Give Reasons:
- 9. Power Level To Which Restricted, If Any (New MWe):
- 10. Reasons For Restrictions, If Any:
This Month Yr-to-Date Cumulative
- 11. Hours in Reporting Period 744 744 8,089
- 12. Number of Hours Reactor Was Critical 596.1 596.1 2,134.7
- 13. Reactor Reserve Shutdown Hours 0 0 0
- 14. Hours Generator On-Line 474.8 474.8 1,318.79
- 15. Unit Reserve Shutdown Hours 0 0 0
- 16. Gross Thermal Energy Generated (MWH) 1,398,655 1,398,655 3,070,685
- 17. Gross Electrical Enersy Generated (MWH) 497,248 497,248 1,054,569
- 18. Net Electrical Energy Generated (MWH) 474,398 474,398 993,238
- 19. Unit Service Factor 63.8 63.8 63.8
- 20. Unit Availability Factor 63.8 63.8 63.8
- 21. Unit Capacity Factor (Using MDC Net) 55.5 55.5 55.5
- 22. Unit Capacity Factor (Using DER Net) 55.5 55'.5 55.5
- 23. Unit Forced Outage Rate ? 7 36
- 24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):
2-18-81 5 Days Ice Cond. Ice Basket Weighing 4-01-81 Low Voltage Mitigation Proble=s
- 25. If Shut Down At End Of Report Period, Estimated Date of Startup:
- 26. Units In Test Status (Prior to Commercial Operation):
Forecast Achieved INITIAL CRITICALITY 7-04-80 7-05-80 INITIAL ELECTRICITY 8-21-80 7-22-80 COMMERCIAL OPERATION 2-21-81 (9/77)
UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-327 UNIT NAME Sequoyah I DATE 2-2-81 COMPLETED BY Michael G. Eddings REPORT NONTH January, 1981 TELEPHONE (615) 842-0295
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Se em n o ao 3 Licensee U Cause & Corrective No. Date % .S 0 8 .o .S % Event @'i E*e Action to gg g jUj Report # #]
bQ 85 g]* Prevent Recurrence o
& ti &* o Emfo a 30 1-03-81 F 14.4 G 3 N/A
- Reactor Trip 4 1-11-81 S 27.9 B 1 N/A Power Reduction to 57% for Maintenance on Feedwater Pump 31 1-15-81 F 11.5 A 3 N/A
- Rx Trip 32 1-18-81 F 2*5 A 3 N/A
- Rx Trip 33 1-18-81 F 4.05 A 3 N/A
- Rx Trip 34 1-19-81 F 164.1 A 2 N/A
- Rx Trip 35 1-26-81 F 44.36 A 3 N/A'
- Rx Trip 36 1-28-81 F 6.7 A 3 N/A
- Rx Trip
- See_ Attached._Shee.t for caus_ca_k_ Corrective Actions.
1 2 3 4 F: Forced Reason: Method: E::hibi t G-Ins t ructions S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data B-Haintenance or Test 2-Manual Scram. Entry Si:cets for Licensee C-Refueling 3-Automatic Scram. Event Report (LER) File (NUREG-D-Regulatory Restriction 4-Other (Explain) 0161)
E-Operator Training & License Examination F-Administrative
, G-Operational Error (Explain) 5 (9/77) H-Other (Explain) Exhibit I-Same Source
4 UNIT SHUTDOWNS AND POWER REDUCTIONS Reactor Trip No. Causes Corrective Action
- 30 When power was restored to the Instructions were revised for generator / transformer relay re powering the generator /trans-auxiliaries, the EHC system former relay auxiliaries.
switched to load control at the minimum load setting causing an E 81% load rejection. This caused a Lo-Lo steam generator level condition that tripped the reactor.
- 31 The 30 second delay on the 180-82 Electrical maintenance replaced relay malfunctioned when a low flow relay 180-82 and verified the alarm was received on the low voltage delay time.
bushing oil pump causing a generator trip and subsequent reactor trip.
- 32 Oscillations occurred in the EHC con- Oscillations are still under trol system governor valve position investigation.
limiters. The governor valves closed, steam generator levels dropped re-sulting in the reactor trip.
- 33 Reactor tripped due to a Lo-Lo steam N/A generator level in S/G #4 during recovery.
- 34 Manually tripped by UO due to oscil- Oscillations are still under lations in the EHC system and other investigations.
plant instruments.
- 35 Reactor tripped due to Lo-Lo steam Air pressure was increased to generator level in S/G #3 because the feedwater level control of slow response on one of the feed- valve.
water level control valves.
- 36 1-FCV-1-111 failed open causing a 1-FCV-1-111 was repaired.
Hi-High level on S/G #2 causing a turbine trip and subsequent reactor trip.
Plant Maintenance Summary The following significant maintenance items were completed during the month of January, 1981:
- 1. Installed stainless steel casing and rebuilt 2B-B centrifugal charging pump.
- 2. Installed new coated #2 and #3 runners in Unit 2 reactor coolant pump seals.
- 3. Repairs to a printed circuit board were made to vital inverter 2-II.
- 4. Repairs were made to the solid state power supply and firing circuits on the Unit 1 turbo generator exciter.
- 5. A high voltage problem exists that is burning up small motors especially in the radiation monitors. Bad motors are being re-placed with new motors.
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9 Outage Maintenance Summary January, 1981 The following significant maintenance items were completed during the month of January, 1981.
Unit 1
- 1. Work continued on hanger repair and inspection as per IE Bulletin 79-14.
- 2. Reinforcement bars were installed in the personnel airlocks (EL - 690' and 734') to attain a 45 psi yield strength for the airlocks.
- 3. Removed the steam strainers from the #3 and #4 steam lines.
- 4. Removed the #4 coupling cover from the main turbine and the top half end bells on the generator for inspections.
Unit 2
- 1. Installed new Westinghouse seats on the governor valves.
- 2. Work continued on the replacement of the vital inverters.
Unit 0 Or Items Affecting Unit 1 and 2
- 1. Work continued, with part of the system in operation, on the replace-ment of the Public Safety repeater antenna system.
- 2. The re-routing of train B power cables for the fire pumps continued.
- 3. Structural restraints to reinforce the concrete block walls in the auxiliary building to seismically qualify them was completed on the ones that affect the safety of Unit 1 equipment. Work is continuing on the remaining walls.
- 4. Continued working on adding auxiliary control capabilities for the station fire pumps.
- 5. Modified pipe supports and replaced piping to HVAC equipment serving the electric board rooms and main control room.
9 APPENDIX A DOCUMENTATION FOR ODCM CHANGES i
l l
l
. SEQUOYAH NUCLEAR PLAR OFFSITE DOSE CALCULATION MANUAL Effective Page Listing Page Revision 1 through 5 original 6 through 7 revision 1 8 original 9 revision 1 10 original 11 revision 1 12 through 12a original 13 through 14 revision 1 15- . original 16 through 19 revision 1 Table 1.1 (2 pages) original Table 1.2 original Table 1.3 (8 pages) original Table 1.4 original Table 1.4a revision 1 Tables-1.5 through 1.6 original Table 1.7 revision 1 Table 1.8 original Figures l.1 through 1.3 original 20 through 20a original 21 through 27 revision 2 28 original Table 2.1 (3 pages) original 29 through 30 original Table 3.1_-1 (4 pages) original Table 3.1-2 through 3.1-3 original Table 3.2-1. (3 pages) original Figures 3.1-1 through 3.1-4 original 18 revision 2 i
L . _ .
ravision 2 .
18
- - e h
where:
DTH y4 = thyroid dose from C-14, mrem.
Q 4 = monthly release of C-14, Ci.
= C-14 milk ingestion dose factor, crem/y per Ci/m'(Table 1.7) l1 DF 4
X/Q = relative dispersion factor,1.54x10-6 s/m'.
3.15x10' = s/y.
For determining the total thyroid dose:
]
DTH131 + DTH14 (1.28)
DIH = 0.9 where: .
DTH = thyroid dose, mrem.
~
- DTH = thyroid dose from release of I-131, mrem.
737 DTH y4 = thyroid dose from release of C-14. mrem.
0.9 = fraction of total thyroid dose expected to be contributed I
by these radionuclides (actually 0.96).
Step 2 This methodology is to be used if the calculations in step 1 yield doses that exceed applicable limits.
Doses for releases of iodines and particulates shall be calculated using the cethodology in Section 1.1.1, step 1, part B, with the following l
exceptions:
~ ~
1 i
- 1. All ceasured radionuclide releases will be used.
- 2. Dose will be evaluated at real cow locations and will consider actual .
grar.ing information.
3 The receptor having the highest total dose is then used to check
' .. pegg I
compliance with specification 3.11.2.3.
i i
. revision 2 21.
where:
Cg ,= undiluted effluent concentration of radionuclide i, as deter =ined in Section 2.1.2, UCi/nl.
MFC.1
= the MFC of radionuclide i, as specified in Section 2.1.1, UCi/al.
R = the sun of the ratios for release point j.
. 3 There are 4 possible liquid release points into cooling tower blowdown.
- 1. Stea: Generator > ua 4 Cooling Tovar 31ovdown = F
'l I
- 2. Stea: Generator : '
- 3. Condensate Tanks >
l
. 4. Radvas:e Tanks :
" 1
- . v
. The sum of the ratios at the diffuser pipes must be <1 due to the releases from any or all of the above sources. The following relationship will assure this criterion is met:
f (R -1) + f 2 (
1 1 -1) + f (R -1) + f,(R -1) < F 3 3 4 4 - (2.3) where:
f 7,f 3,f ,f; 3 = the effluent flo* rate (;allens/cinute) at t'ne respective release point deter =iend by plant pe risonnel.
1 ,R,,R ,R P = the sus of the ratios cf the respective release
. 3 4 point as deterrined by Iquatica 2.2.
F = =ini=un dilution flev rate f or prerelcase analysis
' coling tower blovicun, gallons /cinute) = 13,000 gait =1n.
I
1 h .
revision 2 21a l
} For releases into the cooling tower blowdown line, additional f
mixing is assumed to occur in the diffuser pond when SON is i l
- operating in Helper or Open cosling mode. ) j s
The results, from field tests conducted in September 1979, are expressed in terms of relative concentration r:
l r = 2.4x10-6 (p + f +f2+f3+f) 4
). ..
j Equation 2.3 then becomes 2.4x10-6 (y + g7 + g2 + f3+f) 4 f 1(Ry -1) + f 2(R2-1) + f 3(R3-1) + !
f 4(R4-1) <F (2.3a) i 4
4 1
I l
s e*
j .
i ,
I rsvision 2 22
, 2.2 Instrument Setpoints
- 2.2.1 Setpoint Dete*mination
- The setpoint for each liquid effluent monitor will be established using J
plant instructions. Concentration, flow rate, dilution, principal gnmna
\
=
emitter, geometry and detector efficiency are combined to give an i
equivalent setpoint in counts per minute (cpm). The physical and technical
, description. location and identification number for each liquid effluent radiation detector is contained in plant documentation. i i
The respective alarm / trip setpoints at cach relcase point will be set such that the, sum 'of the ratios at each point, as calculated by Equation
, 2.2, will not be exceeded. The R is directly related to the total i , concentration calculated by Equation 2.1. An increase in the concentra-tion would indicate an increase in the respective R . A large increase would cause the lictics specified in Section 2.1.1 to be exceeded. The f
minimum alarm / trip setpoint value is equal to the release concentration.
but for case of operation it may be desired that the setpoint(s) be set above the effluent-concentration (C3 ).
That is, .
S 3 = b) x C) (2.4) or bj '= "c -
3-where:
S) = desired alarm / trip setpoint at release point j.
b) = scaling f actor to prevent alarms / trips due't'o variations in the efflue,nt concentrations at release point j.
Cj = total concentration in the liquid effluent at release point j specified by Equation 2.1, UCi/nl.
- l-c -__-
i revision 2 '!
I!
22a :L i
i 1
j' The R used in Equation 2.3a must also be scaled by the corresponding scale
- factor. Equation 2.3 and the corresponding alarm / trip setpoints become l
- ! 2.4x10~b (F + ft+f2+f3 + f4 ) [f (b 1 R-1) t +
1
{ .
f2 (b2 R3 -1) + f3 (b3 R3 -1) + f4 (b4 R4 -1)] 1 F (2.5) j
' o
! . s i
- t b --
t
. 1 C (2.6) !
! 1 i
j 'S '
b = - -
(2.7)
. 2 C
, 2 .
l S '
! b = 3 (2.8) i 3 -
C -
3
' S b4,J (20 i
- C
. 4 e
e r.
I I-e 4
- e empen =
- 4 se
=
0
/ <
P
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _...__-_m_ - . _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ . . _ _ . _ _ . _ _ _ _ _ . _ - _ _ _ _ _ . _ _
D 9 *D'T <
'. . oc cf,]- 3 ravision 2 i<
};
2,3 ];
For example, for 2 release points, minimum dilution flow and no diffuser b:
. pond dilution this becomes, l
S -
g 2 i f
y (13-1 xR) -1+f 2 *
(7-2 x R2 ) -1 ,
< 13,000 (2.10) ,,
i
, 2.2.2 Post-Release Analvsis !
A post-release analysis vill be done using actual release da:a to ,
i
= . ;
ensure that the li=its specified in Section 2.1.1 vere not exceeded. ,
A co=posite list of concentrations (C ), by isotope, vill be used l vith the actual liquid radvaste (f) and dilu:ica (F) flow rates (or volenes) during the release. The data vill be subs:i:u:ed into Equatica 2.3 to !
de=enstrate cc=pliance with the lici:s in Section 2.1.1. This da:a and setpoints vill be recorded in auditable records by plant parsonnel, i
. 2.3 Dose 2.3.1 RETS Recuire ent u
Specification 3.11.1.2 of the Radiological Effluent Technical Specifi- ,
cation (RETS) requires that the' dose or dose co =ittent to an individual ,
frcs radioactive =a:erials in liquid effluen:s released to unres:ricted
- areas from each reactor (see Figure 2.2.1-1) shall be limited:
- a. During any calendar quarter to < 1.5 =re: to the total body and :o l
< 5 =res to any organ, and
- b. During any calendar year to < 3 tre: :c the :ctal bcdy and :c 1 ';
nres to any organ.
To ensure cc:pliance, ce=ulative dose calculatiens will be perferred a:
least once per :onth according to the fc11:ving re:hedology. ,
2.3.2 Monthly Analvsis Principal radionuclides vill be used to conservatively es: ira:e f.;e centhly contribution to the cu=ulative dose. If the projected do's'e exceeds the above limits, the =ethodology in Sectica 2.3.2 vill be i=ple:ented.
?
ravision 2 1
, 24
. The following radionuclides contribute more than 98 percent of the total esticat,cd dose from design source terms. (Table 11.2.7, SQNP FSAR
. source terms). i Percent of Percent of Critical i Total Body Dose Organ Dose (Thyroid)
! . Ingestion Fish Ingestion Fish
, H-3 51.2 .1 13.4 .1 I-131 .4 -
60.2 7.8 I-133 .1 -
12.3 .2 Cs-134 22.4 47.0 5.8 43.2 Cs-137 25.1 52.8 6.5 48.6
~
99.2 99.9 98.2 99.9 A conservative calculation of the =onthly dose will be donc according to the following procedure. First, the monthly operating report containing the release data vill be obtained and the activities released of each of
- the above 5 radionuclides will be noted. This information will then be
, used in the fol' lowing calculations.
1 2.3.2.1 Water Ingestion
. The dose to an individual from ingestion of water is described by the following equation.
i 5 Dj k " . 5 (DCF)g xIik rem (2.11)
_a i=1
,' where:
Djk = dose for the j ch organ from 5 radionuclides from the
[ k th release point rem.
.j = the. organ ol interest' (thyroid or total body) .
~
I l
k = the release point of interest (cooling tower blowdown
~
or turbine building su=p).
.95 .= conservative correction factor, considering only 5 racionuclides.
1
. . i - - - . . ,. .. . .- - -
ravision 2 25 i
DCF = adult ingestion dose commitment factor for the j th organ from the i th radionuclide ren/pCi, see attached as Table 2.1. '
Iik = m nthly activity ingested of the i th radionuclide from the kth release point, uCi.
If is described by-3 'A i Y , DCi 1
Iik = (2.12) 12 u d where:
365 = days per year Aik = activity released of ith radionuclide during the month from the kth release point, pC1.
V = average rate of water consumption (730 ml/d ICRP 23, p. 358) 12 = months per year u = total cooling tower blowdown during releases, ml.
d = diffuser pipe dilution.(5).
i The dose equation then becomes 5
6 Dj k. 4. 67x10 (DCF)ij x Aik mrem (2.13) u l i=1 For continuous releases into the cooling tower blowdown line, additional mixing occurs in the diffuser pond when SQN is operating in Helper or Open cooling mode. The relative concentration r, is given as a function of total flow into the pond:
r = 1.48x10-14 o (2.13a)
Multiplying equation 2.13 by r, the dose from cooling' tower blowdown becomes:
D = 6.906x10-8 (DCF)f3 x Aik "#8 (* }
i=1
ravicion 2 25a 2.3.2.2 Fish Ingestion The dose to an individual from the consumption of fish is described by Ecuation 2.13. In this case the activity ingested of the jth radio-nuclide (Ig) is described by A B M I = ik i , pCi (2.14) ik ud where:
A = a civity released of ith radionuclide during the month k
from the kth release point, pCi 6e O
., revision 2
. 26 ch B = fish concentration factor of i radionuclide , ff ,,
. attached as Table 2.1.
M = a=ount of fish eaten conthly (1.9x10 gm) 3 U = total cooling tower blowdown during releases, ml
. d = dif fuser pipe dilution (5)
The dose equation then becomes - > s,.
. _ 5 4x10 5 Djk
- u A xB x DCF ij mrem (2.15)
. i For continuous releases into the cooling tower blowdown line, additional mixing occurs in the diffuser pond when SQN is operating in Helper or Open cooling mode. The relative concentration r, is given as a fuction of total flow into the pond:
r = 1.48x10-14 u (2.15a)
Multiplying equation 2.15 by r, the dose from cooling tower blowdown becomes: 5
-9 D jk = 5.91x10 Aik
- 31 x DCF fj rem (2.15b) i=1
ravisica 2 26a If these calculated monthly doses exceed limits specified in Section 2.3.1, then a more accurate and complete calculation will be donc as described in Section 2.3.3. An annual check will be made to ensure that the monthly dose estimates account for at least 9.5 percent of the dose calculated by the method described in Section 2.3.3. If less than 95 percent of the dose i
has been estimated, a new list of principal isotopes will be prepared.
2.3.3 Quarteriv and Annual Analvsis A complete dose analysis utilizing the total estimated liquid releases for each calendar quarter will be performed and reported as required in i
Specifications 6.9.1.6 and 6.9.1.8. This analysis will replace previous estimates calculated in Section 2.3.2 and consists of the folicwing approach. [hedosetothej organ from m radionuclides, D ,
is described by ~ "
Dj k = .D ijk, rem (2.16) i=1 m._,
=
I ik, (DCF) ) x rem (2.17) i ='1' oww
\
O l
_. m
.o
.. ..Where:
th
/ D ijk = dose to the jth organ from the i th radionuclide from the k release point, rem. ;
j - the organ of interest (bone, GI tract,' thyroid, or total body).
\
(DCF)ij = adult ingestien dose censittent factor for the,j organ from the i ch radicauclide, renh C1, see Table 2.1.
I g.,- activity ingested of the i th radionuclide from the k th release point, VC1.
i f I f r water ingested is described by
,ik f A V n ik ,uci (2.13)
It=
l %
., and for fish ingestien Iiis described by I
. A B M I = ik i ,pci (," ", c~'3
. fk d
~
whera 6 . -
Ag = activity relased of jth radionuclide duting the release period from the k th release point.wC1.
'. V = average rate of vater consu ; tion (730 =1/d).
n = nu=ber of days during the release ;eried (d).
U = cooling tover,blcrdevn'during the release period, 1.
d = credit for diffuser pipe dilutica (5).
. 'Cif; 3g = fish concenthatien factor of the i " radienuclide, _Ci.n~_
M = anount of fish esten =enchly (1.c 10 x 2
- )
At the end of the year an annual dose analysis will be perfor=ed by calculating the sus of th_e_ quarterly doses to the_ critical receptors.
2.4 C ornility of *_iquid hivasta Ecui- c c
... -Specification 3'.11.1.3 of the h diological Iffluent Technica 5;ecifi-cations requires that the liquid radvaste syste: shall be used to redu e
~
4 o
APPENDIX B DOCUMENTATION FOR MAJOR CHANGE TO RADIOACTIVE WASTE TREATMENT SYSTEM I
l 1
F PORC Review Documentaion This Appendix contains PORC =eeting minutes and USQD with PORC-approved STEAR form document PORC review and approval of the proposed waste treatment change. The minutes of September 8,180, show evidence of PORC review and approval of SQ STEAR-22, Radwaste Handling System.
L b !. -
MINUTESOFMEETINdNO.1366 .
V PLANT OPERATIONS REVIEW COMMITTEE SEQUOYAH NUCLEAR PLANT September 8,1980 This meeting was c9nvened to consider the items listed below.
Those in attendance were:
W. F. Popp, Assistant Power Plant Superintendent C. E. Cantrell, Assistant Power Plant Superintendent J. W. Doty, Powar Plant Maintenance Supervisor (M)
R. L. Hamilton, Quality Assurance Engineer R. J. Kitts, Health Physics Supervisor J. M. McGriff, Jr., Power Plant Results Supervisor D. J. Record, Power Plant Operations Supervisor Visitors : S. Butler, NRC Resident Inspector M. W. Halley, Preoperational Nuclear Engineer W. H. Kinsey, Jr., Asristant Power Plant Results Supervisor
- W. A. Watson, Power Plant Maintenance Supervisor PORC reviewed and recommended approval of the following Preoperational Test Instructions.
V 1. No. TVA-23A (Rev. 0) - Thermal Expansion Of Piping Systems (Main Steam Piping)
, Unit 2.
- 2. No. TVA-34 (Rev. 0) - Nitrogen Supply System - Unit 2.
PORC reviewed and recommended approval of the following Temporary Change Forms.
- 1. Change No. 80-1694 for si-410 (Rev. 5) r antainment Upper, Lower, Incore Instrument Room, Purge, Waste Gas Decay T. Gt Release - Units 1 & 2.
- 2. . Change No. 80-1695 for SU-1.0 (Rev. 3) - Plant Measurements Operational A And Baseline Data - Unit 1.
- 3. Change No. 80-1696 for SOI-30.1 (Rev. 7) - Control Building And Control Room Heating, Air Conditioning And Ventillation System - Unit O.
PORC reviewed and recommended approval for performing SQ STEAR-22.
Page 1 of 3 L
, s b---v 2
MINUTES OF MEETING NO. 1366 SEQUOYAH NUCLEAR PLANT September 8, 1980 PORC reviewed and recommended approval of the SU-7.1 Test Report submitted by the Results Sectica. This startup test established the NSSS startup sequence.
PORC reviewed and recommended approval of the 50-7.2 Test Report submitted by the Results Section. This startup test established several tasks: 1) tak-ing the reactor critical for the first time, 2) establishing the neutron flux 4 range for zeia power physics testing, 3) verifying the proper operation of the reactivity computer, and 4) obtaining overlap data between the source range and intermediate range excore deter, tors.
PORC reviewed and recommended approval of the SU-7.3.1 Test Report submitted by the Results Section. This startup test established the endpoint boron i
concentration, the isothermal temperature coefficient, and the moderator temperature coefficient at various rod configurations.
PORC reviewed and recommended approval of the SU-7.3.2 Test Report submitted by the Results Section. This startup test established the neutron flux dis-
\g,,/ tribution at zero power for different control rod configurations.
PORC reviewed and recommended approval of the SU-7.4 Test Report submitted by the Results Section. This startup- test established the integral and differ-ential worth during zero power physics testing of control banks A,B, C, and D.
t PORC reviewed and recommended approval of the SU-7.7 Test Report submitted by the Results Section. This startup test established the measurement of the differ-l ential and integral reactivity worth of shutdown banks D and C using the reacti-vity computer. The test measured the critical boron concentration of all rods
- in, except the most reactive RCCA, F-10, which is left full out. It also measured l the total reactivity worth of all control and shutdown banks minus the worth of the most reactive RCCA, F-10.
(
l Page 2 of 3 I
I \% s/
L l
s, m Mw -
3 MINUTES OF MEETING NO. 1366 SEQUOYAH NUCLEAR PLANT September 8. 1980 The format and content of plant instructions and revisions thereto listed in these minutes are in compliance with plant Quality Assurance requirements.
h O &
(j,LQA Supervisor Approved '1 kp',_.
Cliairman I /
N C:JF cc: ARMS, 640 CST 2-C H. N. Culver, 249A,13B-K J. G. Dewease, 1750 CST 2-C R. M. Maxwell, ROB-H Q L. M. Mills, 400 CST 2-C F. A. Szczepanski, 417 UBB-C Page 3 of 3 t
....- .% o F 2/A t v .. . o s . .. -
~'
UNn ED STATES GOVERNa'"NT ,
2V[CMOTdndMm TENNESSEE VALLEY AUTHORITY L29 800808 86F TO J. M. Ballentine, Power Plant Superintendent, Sequoyah Nuclear Plant FR03t J. G. Dewease, Assistant Director of Nuclear Power (Operations), 1750 CST 2-C .
^7' AUG 2 61980
SUBJECT:
SEQUOYAH NUCLEAR PLANT - UNREVIEWED SAFETY QUESTION DETERMINATION -
MOBILE LIQUID RADWASTE DEMINERALIZATION FACILITY Attached for your review is the Unreviewed Safety Question Determination (USGo) of the mobile liquid radwaste demineralization system. The Chemical Engineering Group has concluded that no unreviewed safety question exists with this activity.
Because this system is not safety relatc.1, it will not be necessary for the Division of Nuclear Power to sign off the STEAR forms (N-0QAM, Section 4.6, Part II). As in the case of the previously submitted USQD for mobile solidification, we believe it will be sufficient for the plant superintendent. and the Plant Operations Review Committee to review the attached USQD and approve the activity. If you require additional information, please contact Ed Hartwig, of the Chemical Engineering Group, extension 2077, Chattanooga.
O
,a n
'J. G. Dewease RAS:FEH:SSS ~
4 Attachment cc: ARMS, 823 EB-C ^
.h, c,ged DE PG $UPT.
P' PRO -SE 1AH
/ s b E0 b f s/ mean pasa sw, g @,C, u LN.N b MN I
gA*y 6
/
l l (a -
9 7 E 1,
( I
__ L12__c_ D.- _1 D L 1- _ s L_ D ---- J L C - -- b l - - ----
. _. ._, _ ___ m- _ _ - . _ _
,.A 9
UNREVIEWED SAFETY QUESTION DETERMINATION Is the probability of occurrence or the consequences of an accident or malfunction i
of equipment importact to safety previously evaluated in the Final Safety Analysis Report increased? Yes No X Justification The use of a mobile liquid radwaste demineralization system to process liquid radioactive waste cannot impact any of the safety systems evaluated in the FSAR.
The system will be operated in an area which is isolated from any CSSC equipment.
1 Is the possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report created?
Yes No X Justification The FSAR considers accidents which have a potential for serious releases to the environment but it is not intended to cover every routine activity associated with day-to-day plant operations. The FSAR description of the liquid radwaste system includes provisions for spill containment, level monitoring, and continuous air monitoring. There is no analysis of environmental consequences of releases from this system because of the low-activity levels of the radwaste and the low probability that spills from the system could enter the environment. The location and activity levels associated with the mobile liquid radwaste demineralization system afford similar assurances of negligible effects.
l Is the margin of safety as defined in the basis for any technical specification reduced? Yes No X Justification The present Sequoyah Nuclear Plant process control program will be amended as required by Sequoyah Technical Specification 6.13 to allow for operation of the mobile liquid radwaste demineralization system. If all the material contained within the system at any one time were released, the solid and liquid material would be contained in the operation area. The maximum volume of waste which could be released from the system is about 250 ft3 From operational data obtained on similar demineralization systems at other plants, the maximum curie content of this waste would be about 100 curies. Normally the curie content will average 30 curies; however, actual curie content will be dependent upon the activity level and volume of water processed through the system. If, due to a violent rupture, all the activity in the mobile system were released,
'about one' percent of the total activity (maximum of about 0.10 curie) might become airborne. This activity is insignificant when compared to a vaste gas decay tank rupture which could result in an airborne release of 50,000 curies.
A waste gas decar tank rupture has been analyzed in the FSAR from a radiological
- . standpoint and d termined to be an acceptable risk to public health and safety.
FEH:SSS 8/8/80
~ . - , . _ , .. . . , ._ ,..-.-.. . .. - - -
. - - - . . . ~ . - .
~
gm'hvEA u .
'Q'
$}
AtrtfilliENT 1 Page 7 SQA100 t 6/5/80 STEAH COVER S!!EET I? ' . .
kb ' ._ -
Special Test, Experiment, or Activity Request
,. (STEAR) of a I.icenned Unit P PROD Designation STEAR :
}' ..
D.a te : 9/5/80 22 Reference Documents:
9 Originated By: Warren H. Kinsey Administrative tinit J '
of Originator: Sequoyah Results Sect.
Nuclear Plant F. Un'it: Sequoyah Unit 1 Plant System or Feature: Radwaste Handling
- 1. j, :
System Description of and Heason for the Special Test, Experiment, or Activity:
Hookup of a mobile liquid radwaste demineralization facility to the permanent radwaste system.
c.
Yes No Initials Date Does the STEAR affect. any CSSC item? .
Is the STEAR safety related? X pW- 9/5/80 p
- Safety Review for t he S'iEAR bitich Af fects Iten s of the CSSC and are Safety Related.
A Required safety reviews and safety analyses (10 CFR 50.59) have been completed for:
j A. Unreviewed nately epnest ion
! t B. Technical specification limits
- h t EN DES Signolf PEli Signof f P PROD Signoff si p Final Approval for Performing Special Test or Experiment:
ai -
Plant. Operatirenn Review Conanittee Plant Superintendent I
e t 1 C.N.Y Initials f .-. W3/8b 1
,. uite Initta s D.ite
- l, . . - . . .
.s: _ . .
u.
,, . = - - -