ML20062B534
| ML20062B534 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 10/26/1978 |
| From: | Madgett J DAIRYLAND POWER COOPERATIVE |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-09-01, TASK-9-1, TASK-RR LAC-5519, NUDOCS 7810310080 | |
| Download: ML20062B534 (16) | |
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DAllt) LAND POWElt COGI*EftATIVE Ba Grone, 0)'nconan i
$4601 October 26, 1978 In reply, please-refer to LAC-5519 s'
DOCKET NO. 50-409 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.
C.
20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE COILING MATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICEliSE NO. DPR-45 PROPOSED MODIFICATION - SPENT FUEL STORAGE Rcforence:
1)
NRC Letter, Ziemann to Madgett, dated September 28, 1978.
2)
DPC Letter, LAC-5341, Madgett to Director of Nuclear Reactor Regulation, dated June 7, 1978.
Centlenen:
Enclesed with this letter is additional information requested by you in Reference 1 for completion of your review of DPC Technical Report LAC-TR-064, " Environmental Impact Evaluation of Spent Fuel Pool Rack Modification", submitted by Reference 2.
The data required to respond to Question 9 is presently being collected.
A response will be provided shortly in a later sub-mittal.
If there are any questions concerning this submittal, please i
contact us.
1 Very truly yours, i
DAIRYLAND POWER COOPERATIVE John P. Madgett, General Manager i
JPM:NLH:af cc:
(See attached list).
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Director of Nuclear Reactor Regulation LAC-5519 Washington, D. C.
20555 October 26, 1978-cc:
J. Zeppler, Regional Director U.
S.
Nuclear Regulatory Commission Directorate of Regulatory Operations Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Ivan W.
Smith, Esq. Chairman Atomic Safety and Licensing Board Panel U.
S.
Nuclear Regulatory Commission Washington, D.
C.
20555 Mr. Ralph S. Decker Route 4 Box 190D Cambridge, Maryland 21613 Dr. George C. Anderson Department of Oceanography University of Washington Seattle, Washington 98195 O.
S.
Hiestand, Jr.
Attorney at Law
- organ, Lewis & Bockius 1800 M Street, N. W.
Washington, D.
C.
20036 Kevin P.
Gallen Attorney at Law Morgan, Lewis & Bockius 1800 M Street, N. W.
Washington, D. C.
20036 Coulee Region Energy Coalition P. O. Box 1583 La Crosse, Wisconsin 54601 4
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.9RC QUESTION t
1.
Your submittal of June 7, 1978 indicates that the SFP;uill be filled uith 13 feet of water above the storage rache, wherece filling the pool to the maximum pool uater level, at the 700-foot elevation, vould allou an excese of 20 feet above thcee rache.
E= plain uhy this additional 7 feet of uater chielding isn't planned to be used during the modifi-cation to reduce the doce rate levele in operating avece (i.e., over the pact center and edge of the pool) so that occupational e=posures are reduced to levels that are as lov as is reasonably achievable (ALARA ).
DPC RESPONSE Whenever fuel handling is performed in the FESW, the water level i
is maintained at the 700-foot elevation.
During outages and for the proposed modification work, this practice will be followed.
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.7RC OUESTION C.n the collective dose (man-ren) to the' divers be reduced by re-arrangement of the spent fuel, in the S??, so that a greater than 5-foot vater shictd nay result during diver ope.ra tions ?
DPC RESPONSE The effective 5-foot water shield is an estimated value and is considered a minimal value.
Every effort will be made to re-arrange spent fuel assemblies in order to minimize the radiation exposure to the divers.
The addition of lead sheeting under water is also being considered as a means of reducing radiation exposure.
A reliable determination of dose-rate levels and shielding requirements will be obtained after the next refueling.
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!!RC QUE3:Inil 3.
On an ALARA bacia Justify your decision to use modification Plan A considering that Plan B provides a smaller occupa-tional exposure.
DPC RESPONSE Plan B is the preferred plan intended for use and assumes that workers can disconnect all bolts and fittings with the use of tools with extended handles.
In the event that this is not feasible, divers would be required to perform some of the underwater work.
When the June 7, 1978 submittal was issued, a possibility existed a
(and still does) that LACBWR would be shut down due to the inability to discharge spent fuel to the FESW for lack of sufficient storage space.
Thus, Plan A was submitted which would require 100 fewer man-hours to complete.
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2RC QUESTION 4.
Provide an estimate of the dose rate above the spent fuel pool from concentrations of radionuclides in the pool averaging about 1 : 10~3 uCi/nt, as indicated in Section 3.1 of your June 7,
1978 submittal.
DPC RESPONSE The dose rate one meter above the fuel pool water surface from concentrations of radionuclides in the pool averaging about 1 x 10-3 uCi/ml is approximately 2.0 mrem /hr.
The 1 x 10-3 pCi/ml activity represented an early estimate based on a limited number of samples.
It should be noted that based on the average accumulated data for the year 1978 to date, the pool concentration of radionuclides is closer to 3 x 10-3 uCi/ml with a resulting dose rate of approximately 6.0 mren/hr.
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NRC CUESTION 5.
What is the present annual occupational exposure (in man-rem) in the S?? area from all operations in the S???
Describe the impact of the proposed modification on this occupational ex-posure.
Include in your analysis the c pected exposure fron more frequent changing of the demineratiner and filter cart-ridge.
DPC RESPONSE The annual man-rem exposure in the FESW (SFP) area from all oper-ations for the past three years is summarized below:
Man-Rem Exposures Year Refueling All Other Operations (Except Those Described Below) 1978 0.360 3.136 1977 0.893 4.584 1976 0.314 1.655*
- Includes rack modification of 1976.
The proposed modification is not anticipated to have any significant impact on these occupational exposures.
Exposures for other jobs previously addressed elsewhere are:
Routine non-fuel handling exposures - response to Question 13.
Filter changing and demineralizer resin sluicing -
Section 2.2 of LAC-TR-064 (June 7, 1978).
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9 NRC QUESTION G.
Provide the estimated volume of contaminated material (e. g.,
egent fuel racko,-seienic restraints) expected to be removed from the spent fuei poci during modification and shipped to a licensed burial site.
DPC RESPONSE The estimated packaged volume of contaminated material removed from the spent fuel pool and shipped to a licensed burial site is eight hundred (800) ft3 4
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NRC QUESTION 7.
Your June 7,
1978 submittat did not address the impact of the proposed pool modification on the radioactive gaseous effluent from the pool to the environment.
Include in your discussion the change in the annual c=posure to the popula-tion from this source of radicactivity.
DPC RESPONSE There will be no change in the annual exposure to the population due to the short-lived gaseous isotopes.
This exposure is related to the number of elements discharged annually and not to the FESW inventory since these gaseous isotopes decay to neglib-ible quantities after approximately 100 days.
The annual exposure increase per spent fuel assembly to the low population zone due to 85Kr based on a conservative fuel cladding failure of 10% is estimated to be 0.034 mrem beta dose and 0.035 mrem gamma dose.
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NRC OUESTION 8.
Provide the failed fuel fraction for each year La Crosse has operated.
DPC RESPONSE The " failed fuel fraction" is taken to mean the fraction of fuel assemblies determined by the dry sipping process to have cladding defects.
It is not possible to determine this on a year-by-year basis and the information available (cycle-by-cycle basis) is outlined in the following:
Cycle Dates Elements Fraction of Total Cycle 1
- Initial Startup 8*
.111 August 19, 1972 Cycle lA October 14, 1972-20
.277 March 30, 1973 Cycle 2 June 25, 1973-23
.319 November 3, 1973 Cycle 3 December 21, 1973-10
.138 May 9, 1975 4
Cycle 4 August 11, 1975-26
.361 May 11, 1977
- Cycle 1 was concluded prior to the installation of out-of-core sipping equipment.
These elements were removed based on indi-cations of the in-core failed fuel element detection system.
Two of the eight elements were later determined to be satis-factory for reuse.
.7E C '? UES TION 3.
Provide a discussion of the impact of the poci modification on pool leakage.
Include in your discussion the pool ledage at different heights of ucter in the pool, the leakage expected af ter the pool modification, and the capability of the rad-waste systen to process this uater.
DPC RESPONSE The data required to respond to the above request is presently being cc11ected.
A response will be provided shortly in a later submittal.
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.7RC' CUESTIO,7 10.
Discuas the apent fuel pool water level and vater temper-ature inctrumentation. Include the-capability of the instrumentation to alarm and the location of the alarme.
DPC RESPONSE k
The low level alarm fcr the spent fuel pool is installed at the 4
- 680 foot level and will actuate at three inches below the 680 foot level.
The top of the upper tier of racks is nominally'at 677'-9-3/4".
T' tis means 20 fcot-7 inches or 18,000 gallons of i
water would-exist.in the fuel pool at the low letel alarm point.
The low level alarm is brought to the Control Room Annunciator Panel and will give a visua.1 and audible annunciation.
The spent fuel pool-
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water inlet and outlet-temperature is n.onitored and recorded in the Control Room.
This-temperature will cause an alarm in'the Control Room if the temperature exceeds 1350F and will annunciate in the Control _ Room giving a visual and audible annunciation.
The temper-
. ature is also logged by an Operator each 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> on a Control Roon I.og Sheet.
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Two 0-200 F temperature indicators are provided at the fuel storage j
cooler inlet and outlet for local monitoring.
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1, MAY 23, 1978 1
1 TO:
DISTRIBUTION FROM:
ROBERT PRINCE, LACBWR RADIATION PROTECTION ENGINEER-i
SUBJECT:
P2 PORT
" ENVIRONMENTAL IMPACT EVALUATION OF-SPE!;T FUEL POOL RACK MODIFICATION" i
Please note the corrections on pages 2, 3 and B-3 of the subject report.
The values used were for half value layers (HVL) when tenth value layers (TVL) were actually being considered.
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'3RC CUEST Cll 11.
In'Appendi: 3 "In-Plant Radiological Assesament" on page B-3,
'the evaluation of dose rate found from the given mathematical model is 3.24 107 R/hr.
To reduce this value to 30 mrem /hr-requires a factor of reduction of 109 or 9 tenth value layers (TVL's).
The report states that 9 half-value layers (H VL 's ) or r 3 feet of vater is required to provide this factor of reduction.
Since 9 H7L 's only give a factor of reduction of E12, please calcuicte the additional thickness of vater required to provide the appropriate factor of re-9
-duction (i.e.,
10 ).. Include buildup factors in the cai-culation, and relevant referencca.
CPC RESPONSE This question has previously been answered in the revised June 7, 1978 submittal.
Please refer to the attached nemo concerning the-original calculations.
It should be noted that buildup factors incorporated in equation 2 on page B-2 of the submittal.-
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URC QUESTIO.V 12.
When tuo tiere of racks are inctallcd, the top of the second rack vill be at about' the 620 ' level.
Therefore, uhen a fuct element is transferred to thia upper rack, only appro:inately lo feet of cater shielding vill be available above the top of thic ascenbly during the transfer.
What additional dose vill personnel receive,, as a result of this fuel transfer, compared to the lo to 20 feet of shielding that vould be available if a fuel transfer uas rade uith only a single tier involvedi
..DPC RESPONSE SThe - top of Ehe upper tier is at the 678' ' level.
This allows a water depth of.14 feet above the top of the fuel assembly even during the transfer process.
This water depth.is greater than what has been used for fuel transfer in the past.
In order to transfer fuel from the reactor vessel to the FESW, the fuel assembly must clear the bottom of the transfer canal.
Thus, the depth of the transfer canal determines the effective water shield (12') during transfer and not the height of the storage racks.
Since the same amount of water shield will be available, no additional dose to:
personnel is expected during fuel transfer operations.
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- .*RC OUESTION 13.
For the tuo tier design, justify why the cater level in the apent fuel pool vould be belco the 700-foot elevation and why ti:e reculting occupational c:posure oculd be AURA.
In yc:e dia:ucaicn, tabulate when the vater level is belco this elevation, uhy it is belev this elevation, uhat the dose rate could be to personnel at this elevaticn, uhat the eapected cecupany is at this elevation and :: hat is the estimated collective dose (man-rem) during the time when the t. uter level is belou this elevaticn.
In addition, justify chy specifica-tion 4.2.8.3 of the La Crosse Technical Specification 2 should not be a~cnded to rcquire that the rinimic, t.nter levc!, during handling i
and storage of irmdiated fuel assemblies, he near the 700' clevaticn.
DPC RESPONSE During all fuel handling operations, the Fuel Element Storage Well (FESW) water level is maintained at the 700 foot elevation. While the plant is operating, the water level is usually maintained below the 700 foot elevation to decrease the liner leak rate.
The dose rate in the immediate vicinity and/or directly over the pool is 10-30 mrem /hr when the water level is at the 680 foot elevation.
Approximately three feet from the edge of the pool, all dose rate levels are less than 10 mrem /hr.
During plant operation, occupancy times in the vicinity of the l
FESU are minimal.
The FESW is not located in a high traffic area.
Operator tours represent the most significant occupany factors for this area.
These occupancy times amount to approximately two hoursper week on the 701 level.
It is conservatively estimated that, out of this time, 10 minutes per week is spent in the immediate-vicinity of the FESW.
The annual exposure due to this occupancy is 0.17 man-rem based on an average exposure of 20 mrem /hr.
Based on this information, it is folt that the important require-ment is to maintain a minimum water depth over the spent fuel assem-blies and that amended technical specifications requiring that the water level be maintained at the 700 foot elevation is not mandated.
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