ML19269D311
| ML19269D311 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/29/1979 |
| From: | Stallings C VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Parr O Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7906010237 | |
| Download: ML19269D311 (4) | |
Text
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3 VIItu r N I A Er.iccrrenc AN n l'ow ncle COhll%NY it rcumino,vinoin er unuun May 29, 1979 Mr. Harold R. Denton, Director Serial No. 394 Office of Nuclear Reactor Regulation P0/SPS;pmd Attn:
Mr. Olan D. Parr, Chief Docket Nos. 50-338 Light Water Reactors Branch No. 3 50-339 Division of Project Management License No. NPF-4 U. S. Nuclear Regulatory Commission CPPR-78 Washington, D. C. 20555
Dear Mr. Denton:
In response to Mr. Parr's letter of March 23, 1979, the following information is provided:
(i) Amendment No. 66 dated March 23, 1979 revises the FSAR for North Anna Unit Nos. I and 2 to correctly describe the liquid waste deminer-alization system currently being used to meet applicable regulations and the Appendix B Environmental Technical Specifications.etached to License NPF-4.
(ii) Regarding the determination that the originally installed 6 CPM 11guld waste evaporator is inadequate to process the volumes of waste encountered at the station, several factors may be cited which have contributed to this situation.
(a) A basic change in operating philosophy for the liquid waste dis-posal system has led to significant increases in'the volumes of liquid waste which receive processing. The system design originally called for segregation of " clean" wastes flowing to the high level waste drain tanks, according to influent activity level. This design allowed the routing of lower activity wastes directly to the low level tanks, thus bypassing the waste evapo-rator. However, the required sampling and valve line-up proce-dures to accomplish this segregation would contribute to additional operator radiation exposure.
The capability to process these wastes regardless of activity level was thus comisidered to be advantageous; both in lowering occupational exposure to personnel and in effecting further reductions in the activity of liquid effluents.
(b) The Clarifier system was originally designed to treat liquid waste based on its ability to remove the high suspended solids 7906010 234 /
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u vn.um tuctmc asu Pown comxv ro Mr. O rold R. Denton, Director and phosphates contained in steam generator blowdown used for secondary chemistry control. Full steam generator blowdown was assumed to be discharged through the clarifier, thus maintaining the stable inlet water chemistry necessary for proper clarifier operation. With the change to All Volatile Treatment (AVT) within the secondary system, additional equipment and operator control would have been required to adequately filter the non-phosphate sludge produced. Also, the steam generator blow-down system was modified to allow recovery of this water via the condensate polishers. This change drastically affec'ted the chemical stability of waste flows.to the clarifier and resulted in operating problems with the system.
In order.to maintain releases of radioactivity as low. reasonably achieveable (ALARA),
it was necessary to provide alternate treatment for all " clean" wastes.
(c) Construction activities for North Anna Unit No. 2 and terrestrial water inleakage have significantly increased the volume of liquid waste flowing to the auxiliary building sump system.
These inputs were not considered in the original design of the waste disposal system, and have contributed to the current.need for increased treatment capacity.
Although the cessation of ~ conatruction activities for Unit No. 2 and building improvements to reduce terrestrial water inleakage would minimize these contributions to liquid waste, the situations described in (a) and (b) will continue to require the processing of waste volumes beyond the capacity of the liquid waste evaporator.
The operation of the installed liquid waste demineralization system therefore remains the best available alternative for maintaining liquid effluent. releases ALARA.
(iii)
Pre-operational testing of the Radioactive Waste Solidification System has shown that excessive quantities of '? free water" remain following solidification of wet wastes. This free water, which during testing approached 10% by volume, requires additional handling of radioactive waste and thereby results in increases in occupational. radiation expo-sure. The system supplier has not.yet provided information to indicate when a satisfactory solution to this problem can be implemented. Due to this situation, solidification of' radioactive wet wastes, including spent resins, is not considered justifiable from an economic or an ALARA standpoint. Furthermore, dewatered resins are currently con-sidered an acceptable waste form for shipment and burial under the conditions of the Radioactive Material License No. 097 issued to Cham-Nuclear Systems, Inc. by the State of South Carolina. All dewatered resin shipments from North Anna Power Station are currently received by Chem-Nuclear for disposal at the Barnwell, S. C. site.
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vimam tu.crmc eu Powra Cowwv To Mr. Harold R. Denton, Director We are aware of no current Federal regulations which prohibit this means of disposal.
(iv) An evaluation was recently performed to provide a comparison of liquid effluent releases based on current operation of the Liquid Waste Demineralization System vs. use of the Liquid Waste Evapora-tor.
The results of this evaluation have been compared with similar data presented in Table 11.1 of Supplement No. 2 to the Staff's Safety Evaluation Report (SER), to determine the extent to which modifications to the liquid waste disposal system have influenced the characteristics of the effluents (i.e. isotopic dis-tribution). The following table provides a summary of this data for radionuclides comprising a major portion of the liquid waste source terms.
COMPARISON OF~ CALCULATED LIQUID EFFLUENT SOURCE TERMS'FOR NORTH ANNA UNIT NOS.1 AND 2 Curies Per Year Per Unit and Percent of Total (Excluding Tritium)
SAFETY EVALUATION WASTE DEMINERALIZATION REPORT EVAPORATOR SYSTEM EUCLIDE Ci/Yr.-Unit Ci/Yr.-Unit Ci/Yr.-Unit Mo - 99 1.2
(-2) 3 2.7
(-3)
<1 1.5
(-2) 2 TC - 99m 9.8
(-3) 2 2.5
(-3)
<1 1.3
(-2) 2 I -131 1
(-1) 24 4.4
(-1) 70 4.2
(-1) 59 I -133 1.5
(-1) 37 5.4
(-2) 9 5.5
(-2) 8 I -135 7.7
(-2) 19 7.2
(-3) 1 8.9
(-3) 1 Te -132 3.1
(-3) 1 9.6
(-4)
<1 5.0
(-3) 1 Cs -134 4.7
(-3) 1 3.2
(-2) 5 5.3
(-2) 7 Cs -137 3.6
(-3) 1 3.8
(-2) 6 5.3
(-2) 7 Ba -137m 3.1
(-3) 1 1.3
(-2) 2 2.7
(-2) 4 Total 4.1
(-1) 6.3
(-1) 7.1
(-1)
(all nuclides except tritium)
Tritium 5.8 (F2) 5.7
(+2) 5.7 0F2)
As can. be seen, some variation exists in the calculated isotopic distributions for the three evaluations performed.
Differences in the general plant input parameters supplied to the model used for the calculations have been found to be partially responsible.
However, the major difference between the data presented in the Staff's Safety Evaluation Report, Supplement No. 2 and our most recent calculations for the waste evaporator and liquid waste demineralization systems appears to stem 'from the assumed use or non-use of the Clarifier system. The decontamination factors 2266 ;58
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\\ lhUlNIA EttcTunc AND POWE N commy To Mr. Harold R. Denton, Director (DF's) provided by the clarifier system as described in Section 11 of the FSAR, are 10 for iodine, molybdenum and tellurium and 100 for all other isotopes. An additicnal DF of 10 for all isotopes except yttrium, molybdenum and cesium can be assumed if the clarifier's downstream mixed-bed demineralizer is included in the evaluation. Although it is apparent that the staff model included the clarifier, we have been unable to determine the exact clarifier system DF values assumed in their evaluation. Our evaluations do not include the clarifier.
Since this change represents the only major difference in the Staff's calculation and the " waste evaporator" case which we performed, it can be reasonably concluded that this is the major cause for the observed variations. The modification of the liquid waste disposal system to substitute the demineralization system for the waste evaporator has not resulted in significant changes to the isotopic distribution of the liquid ef fluents.
An assessment of the radiation doses resulting from the release of liquid effluents based on current operation of the liquid waste disposal system has been performed using the source term data summarized herein. This assessment confirms that the design objectives of Appendix I to 10 CFR 51 are satisfied.
It is planned to perform the modifications to the liquid waste demineral-ization system necessary to make this equipment permanent; however, a schedule for this work has not yet been established.
If we can be of further assistance in this matter, please contact us.
Very truly yours, I
hd C.M.Stalikgs Vice President - Power Supply and Production Operations cc:
Mr. James P. O'Reilly, Director Office of Inspection and Enforcement Region II e
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