ML20082E929

Provides Documentation & Data Discussed W/Nrc During 910603-06 Meeting & Interface Meeting W/Region I on 910516- 17.Viewgraphs Encl

ML20082E929
Person / Time
Site:	West Valley Demonstration Project
Issue date:	07/23/1991
From:	Rowland T ENERGY, DEPT. OF
To:	Hurt R NRC
References
REF-PROJ-M-32 NUDOCS 9108020141
Download: ML20082E929 (116)

Text

_ _ _ _ _ _ _ _ _ _ _ - - _ _ - - _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .._ _ __. _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

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[#p)%4 Department of Energy Idaho Operations Offico g [>p([_.

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• West Valley Project Office kg[,5 P.O. Dox 191 I

West Valley, NY 14171 J

July 23, 1991  ;

Mr. R. Davis Hurt U. S. thiclear Regulatory Commission ,

Headquarters Washington, D. C. 20555

SUBJECT:

Responses to NRC Comments and Actions from the Interface Meeting of June 3-6, 1991

REFERENCES:

1. Letter, J. C. Cwynar to T. J. Rowland, WD:91:0580, " Response to Action Items from NRC Review Meeting of April 2-4, 1991," dated May 30, 1991

2. Letter, P. S. Klanian to T. J. Rowland, WD 91 0596, " Response to NRC Comments on Coment Recipe," dr.ted May 31, 1991

3. Insert for the Waste Qualification Notebook -

WVHS-TP-025, Rev. 0

4. Letten, J. L. Mahoney to T. J. Rowland, I WD:91 0725, " Meeting Minutes for Coment Waste Form Discussion, NRC Site Visit of June 6, 1991," dated July 5, 1991

5. Selected Well Level and Precipitation /

Temperature Data requested by NRC (Tom Nicolsen)

Dear Mr. Hurt:

This transmittal officially provides to you documentation and data discussed with NRC representatives during both the subject meeting and an earlier interface with your Region I Office on May 16-17, 1991.

e Reference 1 provides information on the-Evaporator Acid Flush

( effort, including split sample" scope which takes into account statistical significance; a tentative schedule for flushing and sampling; and finally information on the accountability of the fissile material in the system.

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R. Hurt July 23, 1991

! Ret. -

sco 2 provides resolution of NRC comments received on Ceme. Recipe Qualification in your transmittal of April 22, 1991. The West Valley Nuclear Services Company, Inc. (WVNS) responses were reviewed with Mary Adams during the June 6 meeting with an informal copy provided.

RefGrence 3, WVHS-TP-025 (4 copies), is an insert to the Waste 1

form Qualification " notebook" provided to you by transmittal dated May 14, 1991. A copy was provided to Mary Adams during the subject meeting for convenience.

Reference 4 is cur version of the " meeting minutes" of the Thursday, June 6 discussions on cement qualification. The minutes reflect some basic agreements on handling additional l Information as it is generated. For example, the Test Plan for cement qualification " changes" as time marcnes on. WVHS documents these changes with a " test exception" control system.

During the meeting we discussed the issue of how frequently these changes should be passed on to NRC. Our agreement with Mary Adams and Gary Comfort was that we would incorporate the changes into revised Test Plan documentation as appropriate and  ;

only forward approved final revisions to you. This will allow us to conduct our activities in a " normal" fashion, and will prevent sending information that may seem somewhat insignificant upon receipt.

The minutes also reflect agreement on earlier discussions between NRC and West Valley on the withdrawal of the proposed " Topical Report" for Cement Qualification. All parties are in agreement that the TSRs (Test Summary Reports) as described in the Waste Qualificatior. Notebook, will provide the same information as is conventionally expected in a To.)ical. This letter documents our mutual agreement and authorizes WVNS to delete the Topical Report deliverable. We will proceed to remove this activity from the schedule, which will be reflectec in one of the upcoming schedule updates.

Reference 5 is some additional data on monitoring wells and weather conditions, requested during the presentation of the NDA Groundwater Model. While we applaud the NRC efforts to develop a good scientific groundwater model, serious concerns over the applicability of the model to the " normal" conditions experienced in the NDA enclosure were expressed during the meeting. The concerns are based primarily on the data against which the model l was correlated - most of it was collected during the construction phases of the NDA Interceptor Trench project. This represents a highly unusual set of conditions which would introduce significant amounts of non-typical information upon which the model is based. Hopefully, the enclosed data will help this situation somewhat. Now that the trench is completed, it should continue to stabilize over time, and should allow the groundwater l

. - ___ _ _ ._ _ -_ _ .- _ _ _. _ _ _ _ __ . _ _ _ ._ _. __ ~ . _ . - ._

R. Hurt July 23, 1991 and run-off to approach more " typical" behavior. However, it would be nice to have much more data (say 3-5 years worth) before the model could be expected to provide valid results. As indicated in the presentation, NRC hopes to publish this model in an upcoming NUREG. We would like to request an opportunity to review the " draft" before it is published, to assist in providing a reasonably accurate model.

Should you have any questions, p?. case call A. Yeazel on (716) 942-4780 or Steve Ketola (716) 942-4314.

Sincerely, 78Rowland, J.

F Director N

T.

West Valley ?roject Office Enclosure cc: P. S. Klanian, WVHS P. C. Newson, WVNS WSK:020:91 - 0894:91:08 0700:91:11 0254:91:10 WSK/s1 t

/

( West Valley WD:91:0580

• o en ist hp-Nuclear Services Company RESPONSES: **'tVp' d '*Yo" 18171 0198

/ DW84379 HS'W Incor00 fated f

DW:4382 DW:4383 May 30, 1991 Mr. T. J. Rowland, Director West Valley Project Office U. S. Department of Energy MS DOE P. O. Box 191 West Valley, New York 14171-0191 1

Dear Mr. Rowland:

Attention: J. A.;Yeazel-

SUBJECT:

Response to Action Items from NRC Review Meeting of April 2-4, 1991

References:

1) WD:91:0414. "NRC Requests per Apr-il 4,1991 Closecut Meeting", M. N. Haas to T. J. Rowland dated April 19, 1991.

2) Standard Operating Procedure S0P 15-4, " Sample Operations",

latest revision.

The surpose of this letter is to provide responses to three of the action items due )y May 30, 1991 resulting from the Nuclear Regulatory Commission (NRC) visit of April 2-4, 1991, during which plans for resumption of waste tank farm processing were reviewed. The com)1ete list of action items is documented in reference 1. The action items to )e addressed by this letter are reiterated below along with the West Valley Nuclear Services (WVNS) response.

ACTION No. 1 Solit Samoles

a. of evaporator acid flush (quantity / frequency to be determined) <

Based on discussions with J. Roth of the NRC on May 16, 1991, l the following sampling protocol was agreed to:

1. The spent evaporator acid cleaning solution will be sampled in concentrates tank 50-15A2 after cooling to ambient cell temperature in accordance with the normal WVNS sampling procedura S0P 15-4-(reference 2). This includes an air sparge prior to and during sampling to assure homogeneity.

2. The number of samples shall be in accordance with a statistically derived sample schedule developed by WVNS (see attachmentA). A total of 35 samples-is recommended with the following disposition:

J.

~

SRC4112 NIIn$l. di.cu coSom,on

()'/00 l 9/ ' / /

Mr. T. J. Rowland - Ship to NRC for immediate analysis 4 Analytical and Process Chemistry (A&PC) analyze immediately 4 Hold for future A&PC analyses 10 Hold for future NRC analyses 10 Hold as archive _Z To,tal 35

3. The sample volume shall be approximately 10 ml.

4. WVNS (G. A. Smith /J. P. Jackson) will package and ship NRC samples to:  !

l Robert Oldham i New Brunswick Lab ,

Building D-350 i USD0E i 9800 South Cass Avenue )

Argonne, Illinois 60439-4899

5. The New Brunswick Lab will analyze the NRC samples for total Pu, Pu isotopes, total V, and U isotopes.

6. The A&PC lab will analyze the WVNS samples for total Pu, Pu isotopes, and total U. A&PC lab does not have analytical capability for U isotopes. Consideration is being given to contract with an off-site lab for these analyses. This will be resolved by June 30, 1991 (new commitment).

ACTION NO. 4. Material Balance

• Prior to resumption of operation of the IRTS, WVNS needs to quantitatively reconcile the amount of fissile material downstream of the Supernatant Treatment Syst<#m WVHS has completed a best estimate of the imount of total plutonium deposited in the LWTS evapora*.or based on gross alpha activity data and specific gravity measurements for the evaporator feed and product tanks obtained during IRTS Operations (campaigns 1 through 21). The evaluation includes an analysis of the uncertainty in the calculations. It was concluded that 359 100 grams (two sigma) of total plutonium is present in the evaporator. The fissile portion of the total plutonium (Pu-239 and Pu-241) is computed by dividing the total gran.s of plutonium by 1.23 (aged to 6/1/91) to arrive at 292i81 grams (2 sigma).

Details of the calculation and the uncertainty analysis can be found in attachment B.

A similar estimate for the amount of uranium deposited in the evaporator could not be done because uranium analyses were not routinely performed during operation of the evaporator. A bounding estimate has been prepared using recent analytical results from the 8D-2 waste tank and the waste dispensing vessel.

SRC4112

o Mr. T. J. Rowland . The conservative assumptions of a linear increase of 300 percent j in the supernatant from the original 1986 value, coupled with a loss of 41 percent (assumed from day 1 of use) in the evaporator, leads to the bounding estimate of 1886 grams of U-233/235.

ACTION NO. 5. Acid Flush Schedule The NRC would like advanced notice of when the acid flush is to  !

take place in order to have the opportunity to be present as well l as an information copy of the LWTS-SAR review.

The NRC will be notified one week in advance of the start of the acid flush of the evaporator which is currently forecast for the week of June 17, 1991. An information copy of the-letter requesting approval of LWTS SAR 005,'Rev. 2 is included as attachment C for transmittal to the NRC. This letter also includes resolution of DOE comments.

This letter satisfies commitments DW:4379 DW:4382, and DW:4383. If you have any further questions or comments relative to the responses provided, please contact P. S. Klanian at extension 4382 or J. C. Cwynar at extension 4283.

Very truly yours,

4. c .L J.UC. Cwynar,UManager IRTS Process Control Engineering

, West Valley Nuclear Services Co., Inc.

DC:91:0055 JCC:src Attachments:

A) Letter CJ:91:0044, "LWTS Evaporator Cleaning - Sampling Schedule",

J. L. Mahoney to P. J. Valenti dated May 22, 1991.

B) Letter CJ:91:0047, "Best Estimate and Uncertainty of Pu in the LWTS Evaporator", J. L. Mahoney to P. J. Valenti dated May 28, 1991.

C) Letter WD:91:0555 (FA:91:0049, RS:91:0024), " Request for Approval of LWTS SAR-005, Rev. 2, Draft C", J. L. Knabenschuh to T. J. Rowland dated May 24, 1991.

cc: W. S. Ketola, DOE-WV Project Office, MS-DOE C. B.-Leek, DOE-WV Project Office, MS-DOE R. B. Provencher, DOE-WV Project Office,-MS DOE J. A. Yeazel, DOE-WV Project Office, MS-DOE SRC4112

- -. 4 s - -. - . _ _ - . __ __._____._______ _ _ _ _ _ ._ _ _ ____. -_____.__..__._._____________u_ _-.---_.,..--<~..._----m-m. - - . - ,

DC:91:0055 ATTAotDir A ,

May 29, 1991

/

3 @ 3 C' 4/ MN

[ IRTS Engineering  !"! g g g jggj CJt91 0044 ,

j May 22, 1991 LWTS Evaporator Cleaning - Sampling Schedula l l

b P. J. Valenti MS-W C D. C. Meess MS-Z26 R. F. Gessner MS-201 MRC (original) MS-MRC P. S. Klanian MS-I R. A. Humphrey MS-42 R. Keel MS-Z-26 D. K. Ploetz MS-305 i This memo documents a sampling schedule for spent cleaning solutions from the LWTS evaporator. The schedule is based on standard

• deviations demonstrated in Waste Dispensing Vessel samples that exhibited a small quantity of solids. A total of 35 samples need to be taken. Four of the samples can be shipped immediately to the NRC for analyses. Four other samples shall be immediately analyzed by the WVNS laboratory. If the percent uncertainty of the mean at the 90% confidence level (equals the standard deviation divided by the square root of the number of samples analyzed times a t statistic) is not less than 10% of the average analysis, than additional samples from the remaining 27 shall be analyzed.

DISCUSSION once the acid cleaning of the LWTS evaporator has startod, samples of the cooled cleaning solution will be taken from the downstream tank 5D-15A1/A2. The samples will provide the data to quantify the amount of plutonium and uranium removed from the evaporator.

If the cleaning solution is a perfect salt solution only a few samples will be needed to adequately determine the plutonium and uranium content. If solids are present then .

typically many more samples are needed to reduce the uncertainty in the fissile content in the . leaning solution.

0001JLM

• DC:91:0055 ATTAcitENT A May 29, 1991 During the investigation into the evaporator fissile accumulation, two samples of the Waste Dispensing Vessel were taken. A small amount of solids were detected. On analysis, strontium and plutonium were found to be tied up in the solids.

This was confirmed by filtration. Before filtration, the two samples exhibited a standard deviation of about 30% relative. In a sample with solids, this is typical.

Enough samples of the ovapor7 tor cleaning solution need to be taken to be certain of the jiutonium and uranium content recoved from the evaporator. If we assume that some small solids will be present, the standard deviation can be as high as 30% relative.

With repeats in the number of samples analyzed, the standard deviation can be expected to fall to 10 - 20% relative.

1 For accountability, we typically want to know to within 10% (90% )

confidence), the mass of fissile material moved from tank to  !

tank. For a small number of samples (less than 60), a Student's t statistic must be used, instead of the commut "2", as in "2 sigma". Table 1 lists the t statistic at the 90% confidence level versus the number of sauples analyzed. The table also shows the percent uncertainty of the mean for two different assumed sample standard deviations.

If the cleaning sclution is essentially just a salt solution, the standard deviation can be expected to be near the 5% mark. The third column in Table 1 lists the percent uncertainty in the mean for the different number of samples analyzed. At four sample 4, the percent uncertainty is under 10% relative to the mean. As shown in the fourth column, when solids are involved in the sample, the number of analyses needed to reduce thw uncertainty

• o under 10% is much higher. At 14 samples, with a sample

,andard deviation of 20%, the uncertainty in the mean is finally under the 10% mark.

These calculations point to a significant number of samples for the spent cleaning solution. The total number of samples recommended and the dispositions are:

Ship to NRC 4

, A&PC analyze immediately 4 Hold for future A&PC analyses 10 Hold for future NRC analyses 10 Hold as archive 7 Total 35 0001JLM 2

. . DC 91:0055 ATTAOMPRr A buy 29, 1991

-t Four of the samples can be shipped immediately to the NPC for analyses. Four other samples shall be immediately analyzed by the WVNS laboratory. If the percent uncertainty of the mean at the 90% confidence level (equals the standard deviation divided by the square root of the number of samples analyzed times a t statistic) is not less than 10% of the average analysis, then additional samples from the remaining 27 shall be analyzed.

Dire.ct any questions about the sampling schedule to the undersigned.

\/

J. L. Mahoney, Senior Engineer, MS-M IRTs'~ Engineering West Valley Nuclear Services Co., Inc.

l JIM WP 0001JIR 3

• DC:91:0055 A77AOI4Ein A May 29, 1991 l 3

l Table 1 l Sample Calculations of the l Percent Uncertainty of the Mean For Two Different Sample Standard Deviations Number Student's 5% Rel. 20% Rel.  !

of t Sample Sample  !

Samples Statistic Stdev Stdev-l 2 6.314 31.6 126.3 3 2.920 10.3 41.3 4 2.353 6.8 27.2 5 2.132 5.3 21.3 3 6 2.015 4.5 18.0 7 1.943 4.0 15.9 8 1.895 3.6 14.3 9 1.860 3.3 13.2 10 1.833 3.1 12.2 11- 1.812 2.9 11.5 12 1.796 2.7 10.8 13 1.792 2.6 10.3 14 1.771 2.5 9.8 15 1.761 2.4 9.4 ,

16 1.753 2.3 9.1 >

h k

0001JLM 4

DC:91:0055 ATTAO N NT B pt Fby 29, 1991 jb IRTS Engineering

Fon Ext. 4183 MS-M gg 7.9 W Lener s ' CJt91 0047 .

De May 28, 1991 l kWo Best Estimate and Uncertainty of Pu in the LWTS Evaporator b l P. J. Valenti MS-W CC: D. E. Carl MS-I P. S. Klanian MS-I I J. C. Cwynar MS-W D. C. Meess MS-Z26 1 D. J. Fauth MS-56 D. K. Ploetz MS-305 l R. F. Gessner MS-201 C. J. Roberts MS-D MRC (original) MS-MRC This memo presents the best estimate of the total grams of plutonium thought to be present in the LWTS evaporator. Based on gross alpha activity data and specific gravity measurements from the evaporator feed and product tanks, approximately 359 1 100 grama (2 sigma) of plutonium is estimated present in the evaporator.

The conservative assumptions that have guided the SAR update (460 grams of Pu-239/Pu-241) remain valid. During the acid wash phase, the recovered plutonium shall be cospared against 359 1 100 grams. Details of the calculation and the uncertainty handling are provided as attachments to this memo.

Contact the undersigned concerning questions about this estimate.

V l

J. ; honey, San Dr Engineer, MS-M IRTS(Engineering W30t Valley Nuclear Services Co., Inc.

Attachment At Estimate of Pu in the LWTS Evaporator Attachment B: Method of Uncertainty Calculation

DC:91:0055 ATTACittFNT U tisy 29,1991 Attachment A Estimate of Pu in the LWTS Evaporator DISCUSSION To compute the amount of plutonium thought to be inside the LWTS ovaporator, a simple material balance is computed. The outlet Pu mass (or C1) is compared against the inlot mass (or Ci) with the difference rcpresenting accumulation (or loss) of material in the evaporator.

i Duo co the concentration effect that occurs in the evaporator, the inlet concentrations of mass (gm/ml) or activity (Ci/ml) cannot be co pared to the outlet concentrations. The inlet concentrations must be converted to the equivalent outlet conditions (weight % solids and dOncity) before a comparison can be made.

Tho inlet concentrations restated in terms of the outlet conditions will be a Pu (sci /ml) = _(Meas a Pu ) * (Wt% ov_tlet) * (density g outlet)

(gWtt inlet) * (density inlet)

Tho outlet activity concentration is subtracted from the restated inlot Pu activity concentration.

delta = a Pu, (pCi/ml] - a Puo (pCi/ml]

If the delta is zero, no accumulation is occurring in the evaporator.

If the delta is negative, Pu present in the evaporator is being fluched from the evaporator. If the delta is positive, accumulation 10 occurring inside the evaporator. The total masc or curies that rcc in in the evaporator is calculated by raultiplyit.] the delta times tho volume of concentrates that were removed from the evaporator.

To perform the computation with the historical LWTS data, a few more dotoils must be added. To match the drum production, hence the volume of concentrates, averages will be created from the batch runs cpplicable to each campaign. This allows one loss figure to be quoted

, for each campaign. Gross alpha measurements will be used to represent c1pha Pu since the supernate alpha activity is more than 99% due to plutonium. In this text and the accompanying tables and graph, when cipha Pu measurement is discucsed, gross alpha measurement is actually icplied.

Alco, the weight percent solids in the inlet and outlet streams are not directly measured. Laboratory experimentation has shown that the waight percent solids in the salt solutions are related to the d noity. With radioactive samples, a measurement of the density provides sufficient information to compute the weight percent solids via Wt% TDS = [ (density - 1) / 0.00475 )

• 0.8726

. 1 DC:91:0055 ATTAQ MFNF B thy 29,1991 Tho volume of concentrates produced per campaign is computed via the uno of 40 gallons of concentrates per cement drum.

A final conversion from curies of alpha-emitting plutonium to grams of plutonium is needed. The conversion, based on the L. E. Rykken tepical-report which defined-the contents of waste tank 8D-2, after cging the material to a nominal mid-range of the STS operations, is ,

3.68 grams of total Pu per Ci of alpha-emitting Pu. -

Tho equations to compute the grams of Pu that are accumulating in the i ov0porator during a LWTS campaign ares a Pu (yci/ml] = (Mens a Pu,) o (calc Wtt out) * (density out).

(calc Wtt inlet) * (density inlet)

Calc Wtt out = ( (density out - 1) / 0.00475 ) ^ 0.8726

^

Calc Wtt inlet = [ (density inlet - 1) / 0.00475 ) 0.8726 j

delta = a Pu g (#Ci/=1) - Maas a Pu, (yci/ml]

l a Pu (C1) = 3.68

• de.ta

• volume of concentrates ,

In the. above 5 equations, or.ly 4 values are measurements of the process streams (excluding the volume of concentrates). Using ctendard techniques for combining uncertainties of multiple mencurements (see Appendix B), the uncertainty in total grams of Pu ,

d: posited in the evaporator during each campaign can be computed.

Tcble A-1 lists the input activity concentratiors and density 030curements for the 21 1RTS campaigns along with the volumes of concentrates. The restated inlet activity concentration and other calculated values are shown in table A-2. The total grams of Pu

-cxpected in the evaporator is 359 with a two sigma uncertainty of 100 grcrs.

l l

l

DC 91:0055 ATTAardnVP B May 29, 1991 Table A-1 IRTS Paed and Product Tank Measurements 15B 15A 15B 15A Vol campaign Density Density a Pu a Pu Concent Number gm/ml gm/ml #Ci/ml #C1/ml gals l l

1 1.234 1.321 0.069 0.079 16040 1.227 1.324 0.024 0.064 1.140 1.298 0.044 0.044 1.244 1.301 0.066 2 1.137 1.317 0.026 0.053 15280 1.174 1.308 0.038 0.067 1.242 1.298 0.056 0.054 1.245 1.306 0.059 0.079 3 1.198 1.306 0.042 0.072 22560 1.204 1.303 0.026 0.067 1.239 1.308 0.092 0.069 4 1.220 1.318 0.044 0.072 13360 6 1.173 1.320 0.058 0.077 13120 1.145 1.326 0.028 0.054 1.094 1.320 0.029 0.038 7 1.117 1.323 0.022 0.036 23920 1.071 1.319 0.022 0.051 1.078 1.316 0.020 0.037 1.068 1.319 0.035 0.050 8 1.089 1.328 0.028 0.051 27840 1.064 1.322 0.021 0.052 1.082 1.331 0.022 0.051 1.062 1.327 0.014 0.055 1.085 1.328 0.025 0.061 9 1.067 1.325 0.018 0.038 27400 1.079 1.325 0.039 0.067 1.080 1.331 0.019 0.059 1.057 1.333 0.013 0.051 10 1.090 1.326 0.029 0.079 5920 1.038 1.311 0.010 0.059 11 1.056 1.334 0.013 0.064 25680

1.077 1.328 0.036 0.056 1.062 1.331 0.015 0.053 1.085- 1.333 0.029 0.056 i

l' DC:91:005S ATTAatuWP U May 29, 1991 Table A-1 (cont.)

IRTS Feed and Product Tank Measurements ISD 15A 15B 15A Vol Campaign Density Density a Pu o Pu Concent Number gm/mi gm/ml #ci/ml gci/ml gals 12 1.081 1.326 0.026 0.061 25720 1.083 1.321 0.026 0.091 1.077 1.309 0.035 0.046 1.074 1.324 0.025 0.053 13 1.088 1.324 0.022 0.064 20000 1.080 1.321 0.038 0.068 s 1.082 1.314 0.028 0.066 14 1.081 1.326 0.045 0.068 24440 1.093 1.327 0.042 0.066 1.073 1.315 0.041 0.071 15 1.091 1.330 0.029 0.064 73680 1.090 1.327 0.035 0.082 1.081 1.333 0.020 0.072 1.095 1.318 0.039 0.054 16 1.064 1.316 0.043 0.095 27360 1.062 1.317 0.029 0.027 1.060 1.328 0.026 0.063 1.073 1.324 0.028 0.075 1.069 1.334 0.033 0.063 17 1.066 1.331 0.035 0.064 25720 1.068 1.328 0.025 0.070 1.057 1.329 0.027 0.056 1.067 1.333 0.029 0.054 1.060 1.333 0.029 0.056 1.060 1.326 0.030 0.056 18 1.064 1.329 0.027 0.046 21960 1.066 1.331 0.029 0.049 1.065 1.313 0.031 0.055 1.067 1.314 0.036 0.047 19 1.053 1.315 0.040 0.056 27000 1.072 1.306 0.028 0.040 1.076 1.316 0.031 0.059 1.074 1.319 0.042 0.064 1.077 14307 0.039 0.072 1.089 1:302 0.045 0.093 l

DC 91:0055 ATTAatErn B thy 29,1991 Table A-1 (cont.)

IRTS Feed and Product Tank Measurements ISB 15A ISB 15A Vol Campaign Density Density a Pu o Pu Concent Number ga/ml gm/ml pCi/ml scl/ml gals 20 1.057 1.299 0.0253 0.0849 26440 1.071 1.307 0.0522 0.0844 1,057 1.324 0.0327 0.0732 1.053 1.295 0.0315 0.0599 1.065 1.319 0.0381 0.0678 l 1.066 0.0404 1.068 0.0596 1.069 0.0420 1.058 0.0251 1.050 0.0231 1.056 0.0484 1.063 0.0424 1.052 0.0310 21 1.051 1.317 0.0271 0.0503 2280 1.064 0.0323 l

l. ._ _ - - . .. . . - . - . - - - . - - - - -

DC:91:0055 A7TA0ff2(r B thy 29,1991 i

Tablo A-2 IRTS Feed and Product Pu Accountability Values Restated Inlet Outlet Delta Concentr Campgn a Pu a Pu a Pu Vol Delta Uncert Number pC1/ml #Ci/ml pC1/ml gals gm Pu gm Pu 1 0.046 0.058 0.012 16040 2.6 1.6 1 2 0.045 0.063 0.006 15180 1.3 1.3 3 0.053 0.069 0.007 22560 2.1 2.3 4 0.044 0.072 -0.005 13360 -1.0 2.8 6 0.038 0.056 0.036 13120 6.6 1.4 7 0.025 0.044 0.060 23920 20.0 2.2 8 0.022 0.054 0.042 27840 16.4 2.3 9 0.022 0.054 0.050 27400 18.9 2.6 10 0.020 0.069 0.026 5920 2.1 0.8 11 0.023 0.057 0.051 25680 18.1 2.4 12 0.028 0.063 0.054 25720 19.2 2.4 13 0.030 0.066 0.051 20000 14.3 2.2 14 0.042 0.068 0.103 24440 35.1 2.8 15 0.031 0.068 0.047 23680 15.5 2.2 16 0.032 0.065 0.095 27360 36.4 2.4 17 0.029 0.059 0.095 25720 34.2 2.1 18 0.031 0.049 0.104 21960 31.7 2.2 19 0.038 0.064 0.101 27000 38.0 2.2 20 0.038 0.074 0.120 26440 44.2 2.1 21 0.030 0.050 0.'. 1 2280 3.6 0.4 l

DC:91:0055 ATTAQEFNr B Msy 29, 1991 i

Attachment B I

j Method of Uncertainty calculation DISCUSSION C1cssical uncertainty calculations begin with the recognition that true random normal variation in each variable is required. All the variables under study must also be independent. When these conditions aro met, calculation of the uncertainty begins with partial dorivatives of the governing equation.

l For this memo, the governing equations were depicted at the end of I Attachment A. When combined into one global equation, the grams of Pu I lost during every batch ist

_ gas Pu = 3680

• Vol * {Pug * ( (Do -1) /0.00475)^0.8726

• D o

+ ((Di -1) -/0.00475)^0.8726 + D - Puo} ,

where Vols liters of concentrate Pu g : measured alpha Pu (#Ci/ml) in evaporator inlet Puo : measured alpha Pu (#Ci/ml) in evaporator outlet D:g measured density (gn/ml) of evaporator inlet D:o measured density (gr/ml) of evaporator outlet Two major types of measurement errors are considered in this analysis:

crror on measurement of the densities and error in the measurement in tho alpha Pu activity concentrations. The error in the correlation tying density to weight % solids is not included in this analysis.

Tho uncertainty due to the four measurements is thought to dominate tho total uncertainty, allowing this contribution to bt discarded.

Alco, the volume of concentrates is thought to be known so well that ito uncertainty can be excluded in the tntal uncertainty calculation.

For a generic governing equation, the combined uncertainty is created vie partial derivatives. For example if the following equation were uccd F=X*Y-Z tha uncertainty in F is calculated from a,2 , g g 2 * (6F/6X)2 , g y2 ,

g gpf gy)2 + ,zt. (gpfgg)2 where: a, = uncertainty in F (likewise in X, Y, & Z) 6F/6X = partial derivative of F with respect to X (likewise for Y & Z)

In our governing equation four partial derivatives are needed. The

. fully developed equation is too lengthy to be presented here. The noth is left for the interested reader.

~

DC:91:0055 ATTAQUWir B thy 29,1991 Tho uncertainties in the governing equation should theoretically come fr:3 a significant number (>20) of replicate samples & analyses of a typical evaporator batch. All that is available, however, are the d:noity and gross alpha measurements for the individual batch runs in cach campaign. The multiple measurements presented in table A-1 can be used estimate the uncertainty in the gross alpha and dens'ty c Ocurements.

Ao operated, the LWTS evaporator produced batches that could arguably be exact replicates from campaign 1 through campaign 21. The argument for similar batches would suggest that the density differences between batches is controlled by the uncertainty in the measurement. Stated differently, this implies that the true batch-to-batch variance is caall compared to our ability to measure the density. This means the ctendard deviation between batches can be used to estimate the actual concurement uncertainty. The same argument will be extended to nross cipha measurements although the variability in the evaporator loss c:chanism could effect the standard deviation estimate. Table B-1 and Figure B-2 present the standard deviations computed from this collection of data.

Tho true measurement uncertainty can be estimated from the multiple otendard deviation measurements. A mean-value calculation would incorrectly over-estimate the true standard deviation (since standard-d0viations are distributed per a chi-square function). An cpproximation to the true standard deviation is to take the logarithm of the standard deviations, average the log-values, and calculate the value of ten raised to that average (data presented in table B-1).

Tho best estimate in the uncertainties for the density and gross alpha toccurements are 1 0.006 gn/ml and 1 0.008 pCi/ml respectively.

Onco the losses ano uncertainty in the losses for all the campaigns cro computed, the grand total and its uncertainty must be computed.

The total loss of Pu into the evaporator is simply the sum of the individual campaigns. The uncertainty, however, must be computed via tho partial derivatives of a governing equation. In this case, the governing equation is simply:

Total = #1 + #2 + #3 + ... + #20 + #21 Tho partial derivatives are all equal to 1. So the combined total uncartainty is the square root of the sum of the squares of the individual uncertainties.

DC:91:0055 ATTA0WNT B May 29, 1991 Table B-1 Best Estlantes of Standard Deviations for Evaporator Concentrates Density and Gross Alpha Measurements (1) (2) (3) (4) (5)

Gross Campaign Density Alpha Log of Log of Number Std Dev Std Dev Col (2) Col (3) 1 6.013 0.014 -1.873 -1.844 2 0.008 0.012 -2.108 -1.907 3 0.003 0.003 -2.599 -2.593 6 0.003 0.019 -2.460 -1.716 7 0.003 0.008 -2.542 -2.088 l 1

8 0.003 0.004 -2.485 -2.362 I

9 0.004 0.012 -2.385 -1.922 10 0.012 0.011 -1.933 -1.967 l 11 0.003 0.005 -2.577 -2.319 12 0.008 0.020 -2.118 -1.702 13 0.005 0.002 -2.290 -2.745 14 0.007 0.003 -2.177 -2.590 15 0.006 0.012 -2.188 -1.928 16 0.008 0.024 -2.121 -1.611 17 0.003 0.006 -2.548 -2.196 18 0.010 0.004 -2.019 -2.401 19 0.007 0.018 -2.172 -1.746 20 0.012 0.011 -1.903 -1.967 average -2.250 -2.089 10^** 0.006 0.008 i

Best Estimat:e of Standard Deviations  !

Evaporator Concentrates E, m

Density & Gross Alpha Gross Alpha Density 0.0300 0.0150 m3 0.0250 p

0.0125 - ---

6 0.0100 --

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O.0000 0.0000 1 2 3 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 l

Campaign Number o + Gross Alpha Density l

Density Best Est Gross Alpha Best Est if x

m Density Best Est: 0.006 gm/mi e Gross Alpha Best Est: 0.008 uCI/mi n. .,..,..

5

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DC 91:0055 ATTACIMEt(P C thy 29, 1991 l

FA 91:0049 Rs:91:0024 i i

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@$f Valley P O Bos 191 Nuclear Services Company ACTION: *"' V"'O' * ' "' ' ' * ' 0

• lDC0rD0 fated WD 2007 May 24, 1991 MS-D l

Mr. T. J. Rowland, Director West Valley Project Office U.S. Department of Energy MS-DOE P.O. Box 191 ,

West Valley, New York 14171-0191 '

Dear Mr. Rowlands

SUBJECT:

Request for Approval of LWTS SAR-005, Rev. 2, Draft C

References:

1) Letter WD:91:0487, J. L. Knabenschuh to T. J. Rowland,

" Request for Approval of the Revised LWTS (SAR-005)," dated May 3, 1991

2) Letter CBL:018:90-0785 91:10 (DW 91:0341), T. J. Rowland to J. J. Buggy, Jr., " West Valley Project (WVPO) Conunents on Revised LWTS (SAR-005)," dated May 15, 1991 The purpose of this letter is to request DOE approval of the attached Safety Analysis Report for the Liquid Wasta Treatment Systese (WVNS SAR-005, Rev. 2, Draft C) which addresses safety and health protection matters regarding the acid cleaning of the LWTS evaporator. This revision incorporates responses to your comments on Draft B which were summarized in Reference 2. Only the pages which have been revised since Rev. I are included as Attachment A.

Attachment 8 susunarizes the resolution of each of the fourteen items of Reference 2. Attachment C is a copy of Reference 2 showing the identification by number of each of the fourteen coaunents. These identification numbers appear in the margin of Attachment A to indicate where the most recent changes have been made (revising Draft B to Draft C.)

DOE approval is requested by May 31 so that the acid cleaning may proceed on schedule. Upon receipt of DOE approval, WVNS will release Revision 2 of the subject SAR.

RLW4300iSEA-163 A $vbsdaary of htinghevas fl0Cttet C0rp0f304

,. DC 91:0055 ATTAQPFRP C May 29, 1991 Mr. T. J. Rowland If you have any questions, please contact Joe Johnson at Extension 4064.

Very truly yours,

/ ',

John L. ha'aenschuh

, Vice Freeinen, e and Manager Environmental Safety, Health and Quality Assurance West Valley Nuclear Services Co., Inc.

FA 91:0049 Rs 91:0024 RJJarlw Attachmentes A) WVMS-SAR-005, Rev. 2 (revised pages only)

5) Resolution of Items from Reference 2 l C) Letter CBL 018891-0785 91:10 (with itume numbered)

I

cc

I. A. Matthews, DOE-WV, MS-DOE R. B. Provencher, DOE-WV, MS-DOR J. A. Yeasel, DOE-WV, MS-DOE RLW4300 SEA-163 l

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ATTACHMENT A l A'I27m1FNP C TABI2 OF CONTENTS (CONTINUED) l i thy 29,1991 sRcrION PACK H.9.6.4 Rupture of Process Line Between the Process Building and 01-14 Building................................. 71 H 9.6.5 Rupture of the LVIS Pipe Chase and/or the Pipe Chase 14aving the 01-14 Building. . . . . . . . . . . . . . . . . .. 72 H.9.6.6 Process Tank Rupture Outside XC3............................ 73 H.9.6.7 Tank Rupture Inside XC3..................................... 73 H.9.6.8 V0G HEPA Filter Failure...................................... 74 H.9.6.9 V0G HEPA Filter F1re......................................... 75 H.9.6.10 Evaluation of Cons,equences of Anchor Bolt and Veld Failures Associated with LWTS Process Vessels................ 75 11 . 9 . 6 . 1 1 Rupture of LWTS Transfer Pipe and Spill of Evsporator Cl a snin g Solu t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 78 H.9.7 Mi no r Ac c i de n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 79 H,9.8 PoJential for Nuclear Criticality........................... 80 H. 9.8.1 Evaporator Acid Vash........................................ 81 l References for Section H.9,0................ ................ 83 H.10.0 CONDUCT OF 1RTS OPERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84 H.10.1 Organizational Structure..................................... 84 H.10.2 Preoperational Testing and Operation........................ 84 H.10.2.1 Administrative Procedures for Conducting the Test Program.... 84 H.10.2.2 Test Program Description.................................... 84 H.10.2.2.1 Physical Facilities.......................................... 85 H.10.2.2.2 Process Operations.......................................... 85 H.10.2.3 Te s t D i s cus s i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 H.10.3 Training Pro 5 rams............................................ 86 H.10.3.1 I n t r oduc t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 H.10.3.2 LWTS Operator Training Program Outline....................... 88 CAF0778:3RM - vii -

. DC 91:0055 ATTAGMFMr C iby 29, 1991 UOLUME IV ,

LIST OF TABLES H.9.6-4 Dose to the Maximally Exposed off Site Individual for the Significant Radionuclides from the Evaporation of Liquid from Spent Zeolite Rosin to the Floor of XC3 H.9.6-5 Dose to the Maximally Exposed off site Individual for the Significant Radionuclides for a HEPA Filter Failure Accident H.9.6 6 Dose to the Maximally Exposed Off-Site Individual for the Significant Radionuclidae for a HEPA Filter Fire Accident H.9.6 7 Dose to the Maximally Exposed Off Site Individual for ths Significant Radionuclidas Evaporated from a Spill of LWTS Evaporator Cleaning Solution Following Rupture of Transfer Pipe H.9.8-1 Maximum Concentrations of Fissile Material in LVTS

.Eysporator Concentrates H.9.8 2 Maximus Allowable Fissile Material Solution Concentrations for L1.*IS Operation (All Vessels)

H.10.2-1 Outline of Test Plans for Functional and Acceptance Testing of LWTS H.S.2.2-1 Quality Imvels of Important Structures, Components, and Systems Associated with the Liquid Vast.e Treatment System l

l 1

i CAF0778:3RM x-

DC:91:005b ATTACHMENT C May 29, 1991 WNS- SAR-005 Res. 2 Draft C H.2.3 RADIOWGICAL IMPACT FRDN ABNOWfAL OPERATIONS Abnormal operations are events which could occur from malfunctions of systems, operating conditions, or operator error. Abnormal events are only of consequence for those systems in the LVTS which process, control, or confine radioactivity. The abnormal events considered (Section H.9.1) are of little consequence in terms of potential for environmental releases.

H.2.4 ACCIDENTS Seven accidents associated with the operation of the LVTS were analyzed (Section 9.6). Accidents analyzed include ruptures of various process lines, a tank rupture inside and outside XC3, HEPA filter failure, HEPA filter fire for the Vessel Off-Gas (V0G) System and the Main Plant Ventilation System, end a spill of evaporator cleaning sclution following a pipe rupture. Releases are assumed to be direct to atmosphere for the various pipe ruptures, while all other releases are from the Main Plant Stack. Doses from thena accidents to the maximally exposed off-site individual for the two process line ruptures are 21 and 0.86 mram, respectively. The differenes between the two accidents p concerns the type of material being handled and quantity of material released. For the tank rupture outside XC3 (SD15A1 in UPC), the dose is calculated as 2.1 E-03 mram. Analysis of a tank rupture accident inside XC3 (Spent Zeolite Tank, 71-D-007) leads to a predicted dose of 1.3 E-05 mrem to the maximally exposed off-site individual. The HEPA failure and fire accidents are assumed to be conservative since the 8D-2 Tank activity is the starting point for the analysis instead of decontaminated supernatant. The dose to the maximally exposed off-sits. individual for the V0G System HEPA failure is 0.34 mrea while the HEPA fire dose is projected to be an order of magnitude higher at 3.4 mrea. The dose for a failure of or fire in the Main Plant Ventilation HEPAs L considered to be the sane as or less than that for the VOC HEl'As. The dose from a spill of evaporator cleaning solution following the rupture of a transfer pipe is calculated as 120 mrem to the maximally exposed off-site individual and 2.2 rem to the maximally exposed on-site individual.

CAF0778:3RM _ _ _ - _ _ _ _ _ _ _ _ _ _

, DC:91:0055 ATTAQMFNr C May 29, 1991' --

WNS-SAR 005 R=v. 2 Draft C o Spent avsporator cleaning solutions H.4.2 STEDCTURAL AND IntmAmICAL SAFETY CRITERIA POR THE IRTS The majority of INTS components will be installed in existing structures such as IC3, PPC, UI4, and UPC. All existing structures have been analysed for structural and mechanical safety in previous documents (NFS,1962 and WDP, Weis).- Minor modifications of structural materials necessary to accoassodate C ".S piping and enuipment in the Process Building are not considered significant enough to change the overall structural safety in the above mentioned analysis. Furthermore, the hasards associated with a loss of

-structural integrity are minor when LW S is compared to previous reprocessing activities. The only' structure which can be considered as new is the concrete belcw-ground trench whi<:h serves as a pipe chase between Tank 35104 and IRTS. Seismic analysis of the trench was not considered necessary due to the relatively.small quantity of activity present at any one time, and the low off-site dose potential when an accidental spill was considered (Section H.9.6.1.5). -

H.4.3 SAFETT PROTECTION SY3TEMS B.4.3.1 cumunar.

LWS has been designed to meet all applicable engineering codes and standards throughout construction, inctallation, and system operation (WDP,1985b).

Specific information regarding the various types of safety protection systems considered necessary for the operation of-LWTS appear below.

H.4.3.2 FROTECTION BY IEILTIF12 COEFIK'MIET RARETERS AED SYSTEMS The primary confinement barrier for the relatively low concentration of radioactive materials processed by LUTS operations is considered to be the uystem piping, tankage, and associated process vessels. Secondary confinement is afforded by Extraction Cell 3 (XC3) which houses the heart of the LM S system, the high efficiency waste evaporator, and the ion exchange columns.

CAF0778:3RM i

u.4n.vvss o..... u . .u y s, .m WNS-SAR-005 Rev. 2, Draft C l

H.4.3.5 Et.DIGIDCICAL PRDIECTION l i

1 Construction and maintenance activities will be performed in accordance with I

WDP-10 " Radiological Controls Manual." Shield walls, confinement and containment structures as well as administrative controls (procedures, training, OSRs, etc.) will be used as necessary to maintain radiation doses to occupationally exposed personnel AIARA. Protective clothing (anti-C's, respiratory protection) will be worn when required by radiological conditions. In addition, precautions such as system decontamination and flushing are required in the case of necessary contact maintenance.

H.4.3.6 FIRE AND EIPiDSION PROTECTION The LUTS wi73 have fire detection equipment, alarm systems, and suppression systems comuensurate with potential fire hazards associated with LWTS

( operation as determined by the WNS Radiation and Safety group.

H.4.3.7 RADIOACTIVE WASTE HANDLIIEC AND S'It1 RACE 1

l LVTS operations will use a number of existing tanks to temporarily store LWTS feed material and process material in different stages of the operation.

l Storage of these liquid asterials in any one process vessel or tank is considered to be a temporary condition requiring no additional safety analysis.

H.4.3.8 INDUSTRIAL AND CHIMICAL SAFETY l

l 1

The WNS Industrial Hygiene and Safety Manual. WDP Oll, will provide the administrative guidelines and control over industrial and chemical safety in r*Sard to the operation of LWTS. The LWTS process will provide for addition of acid or caustic to feed solutions. Analysis of an accidant where acid and caustic are inadvertently mixed is considered in Section H.9.7. Analyris of an accident where an acid cleaning solution is spilled is considered in section H.9.6.ll.

CAF0778:2.G u n uvss a u u u am e suy ), an WVNS-SAR 005 Rm'. 2, Draft C H.4.4 CIASSIFICATION OF S1TDCItfRES. COMPONENTS. AND SYSTE_MS Safety and service classifications of important structures, components, and systems associated with the LWTS are presented in Table H.4.4 1. h procedures and criteria used to arrive at the designated classificathas are contained in Section H.4.4 of Volume I.

Accident analysia (Section H.9.6) shows that no postulated accident will ,

I result in a dose to the maximally exposed off-site individual approaching 500 mram. Therefore, no items need to be classified as Safety Class A or B.

The heating / ventilation (HV) supports, vessel off gas pipin6, and ventilation instrumentation and alarms are classified as Safety Class C. The HV supports and V0G piping are Class C because they ensure confinement of radioactivity.

The instrumentation and alarms for ventilation systems are considered Class C  ;

because their loss of function could allow unmonitored releases. Likewise, l airborne particulate and area radiation monitors are Class C due to the potential for unmonitored releases should failure occur.

Sump piping, instrumentation, and alarms are Class C for reasons of confinement of radioactive material and because loss of instrumentation and alarms could lead to undetected leakage and potential for inadvertent exposure. All other items are Safety Classified as N.

H.4.5 DECOstfAKINATION AND DECO MISSIOtMJC Specifications for LWTS equipment and components which will contain radioactive material call for stainless steel (types 304, 304L, 316, 316L).

This will allow periodic flushing of the system with strong decontamination agents, thua rssoving the imaximum amount of deposited radioactive material.

As a result, personnel dose will be minimized during the decommissioning effort. It must be reco5nized that the LWTS system would most likely be the last of the WVDP systems to be dismantled due to the key role it plays in waste water management.

CAF0778:3RM UVNS SAR-005 Rev. 2, Draft C 1.

pumped to the zeolite ion exchanger (71 D 003) via Pump 71-k-15. When LWTS construction has been completed, the ion exchanger effluent will be sent to one of two 34,000 litre monitoring tanks. After sampling, the liqui.d will be routed to eidher Lagocn No. 4 or 5 or the Interceptors depending upon radioactivity concentrations (see Section H.6.1.1.8). Prior to completion of this system effluent from 71-D 003 will be recycled or fed to the existing LLWTF (02 Plant) for further processing and release to the environment.

Evaporator distillate is pumped from the distillate surge tank via Pump P-15 to a zeolite ion exchanger which is solely devoted to evaporator distillate processing. This ion e.<t>ugar is equipped with a differential pressure monitor and effluent raatation monitor similar to those discussed in Section H.6.1.1.6 for the filter / ion exchange train. Off spec effluent (>3E-07 pC1/mL gross beta) will also be diverted to Tank SD15B or the Interceptors via the PPC manifold.

After extended periods of high TDS evaporator operation it may be necessary to perform a cleaning operation to remove accumulated solids. The solids g

increase radiation background near the evaporator, lower its boiling capacity and accumulate fissile isotopes in the evaporator scale or sludge in the bottom head. The solids form because of chemical changes to dissolved salts as they are heated. Since these solids will not dissolve in water alone it is g necessary to use up to 2M nitric acid (<12%) te convert them into a soluble form. The analyses in section 9.8.1 indicate that criticality is not a concern during evaporator cleaning. However, approximately one gram per liter j{

of boron (as boric acid) may be added to act as a neutron poison for fissile isotopes which may be present.

The cleaning solution is placed in the evaporator and heated until sampling results reflect limited effectiveness cf further scale dissolution.

Condensate is returned to the evaporator co maintain a constant liquid level. The spent solution containing the dissolved solids is cooled and transferred to a holding tank for sampling. This solution will not be suitable for cementing in CSS because of high activity and cement recipe CAF0778:3RM 29

. DC:91:0055 ATTAQMFNP C May 29, 1991 WVNS-SAR-005 Rav. 2, Draft C l

1 qualification requirements; it will be routed back to the high level vaste tank (8D-2) after pH adjustment. All temporary holding tanks and transfer lines will be flushed and/or sparged.

Airborne concentrations of NOx and nitric acid fumes at the main plant stack release point will be less than 36 ppa and 5 ppb, respectively, over a four hour period (Surn, 1991). These concentrations would be approximately 66% g higher if the V0G scrubber and condenser were to become inoperable.

Atmospheric dispersion will further reduce the concentrations (well below the applicable threshold limit values-time weighted average) at potential on-site and off-rite receptor locations. No impact on the integrity of the HEPA filters is expected at such low concentrations of nitric acid fumes.

The neutralized solution will be returned to Tank 8D-2. Therefore, this process will not result in any additional liquid releases to the environment. The cleaning operation will remove radionuclides held up from other waste streams (mostly decontaminated supernatant and sludge washes) .

Therefore, no cet increase in airborne.radionuclide emissions is anticipated from this operation.

Depending on the accumulation of solids, cleanouts may be performed more than once over the life of the system.

H.6.1.1.3 Concentrataa collection and Transfer Evaporator concentrates leave the evaperator at approximately 109 C. They are cooled to approximately 35*C in the concentrates cooler and pumped to Tank SD15A1 or 5D15A2 for temporary storage. Process control limits for this waste stream are as follows:

o High TDS liquids (in the range of 500 ppa to about 40 w/o depending upon the ionic species) will be concentrated to a maximum of 46 w/o dissolved solids.

CAF0778:3RM 30

a nauvsa- ,u 4,u n.n 2 e ruy a, un j

.WNS-SAR 005 Rov. 2, Draft C REFERFECES MR SECTION H.6.0 l1 American National Standards Institute (ANSI) N13.1 1969 (R1982), Specification I 1

and Performance of On Site Instrumentation for Continuously Monitoring {

Radioactivity In Effluents.

Burn, P., 1991 Memo E0:91:0055 to G. G. Baker, "NYSDEC Stack Release Permit Modification for 1RTS Evaporator Cleaning," dated April 26, 1991. i

)

Saha, A. K. ,1986, Document WNS-PN-003 to Distribution, "I,VTS Process Control l Interlocks," Rev. O, dated March 19, 1986.

WDP,1985, Safety Analysis Report, Volume II, Existing Plant and Operations.

-1 WDP,1985a, Safety Analysis Report, Volume IV, Cement Solidification System.

WDP,1985b, Safety Analysis Report, Valume I-II, Vitrification System.

l l

CAF0778:3RM 42 l

l ,

i

. DC 91:00SS A'PTACIMPRP C tkiy 29, 1991 WVNS-SAR 005 R:v. 2. Dreft C H.9.4 ABBORMAL EVENTS - OFF-CAS TREATMDrf Failures associated with the Vessel Off-cas system could cause greater than normal releases of radioactive material. These failures include development of a leak in the HEPA filter or reduced capacity due to high loading or excessive moisture. These conditions are detected by differential pressure instrumentation across each filter. Additional HEPA filtration downstream of the V0G HEPAs exists to prevent any activity which leaks by the V0G HEPA to be filtered prior to discharge up the main plant stack. Another abnormal event would be the failure of one of the VOC blowers (6K-2 or 2A). Since a duplicate blower arists, the system can be operational at all times and 1 provide for uninterrupted service during filter changeout. Additional abnormal eients for this system have been described in Table B.9.1-2 of Volume II.

H.9.5 ABBOEMAL EVElfr5 - VIIrrIIATION SYSTEM failures associated with the main plant ventilation are similar to those of the VOC system. A HEPA filter ~ develops a leak, or roughing and HEPA filters develop high loadings. Other failures include damper failure or Heating /Vantilation blower failure. Problems with HEPA filtration increase the potential for increased release of activity. Instrumentation exists to detect filter failure or high differantial pressure. Damper failure or blower failure will cause loss of ventilation and possible air reversals. Like the V0G, two ventilation trains exist for the Main Plant Ventilation to ensure uninterrupted service.

H.9.6 ACCIDENTS H.9.6.1 ACCIDENTS ANAIJZED Seven major accidents have been analyzed in this section. They include: 1)

Rupture of a process line between the process building and the 01-14 building;

2) Rupture of a process line in the LWTS pipe chase and/or the pipe chase leaving the 01-14 Building; 3) Two tank rupture scenarios involving 4 tank inside IC3 and one outside XC3; 4) Vessel Off-Gas HEPA filter failure; 5)

Vessel Off-Cas HEPA filter fire; 6) Main ventilation HEPA filter failure or I fire; and 7) Spill of evaporator cleaning solution following a pipe rupture, r1 CAF0778:3RM i

w asvuss m u-o s u .ms, on UVNS-SAR-005 Rev. 2, Draft C The above listed accidents are considered to represent the most i.evere case for procedural or equipment failure in the LUTS. LUTS has been designed to preclude such occurrencea nowever, the probability of occurrence is still considered greater than zero for the above cases.

H 9.6.2 SOURCE Time The source terms for the above spill-related accidents are highly dependant upon where in the process sequence the event occura. Evaporator cleaning solution will have a relatively higher concentration of uranium and plutonium (up to 120 pCi/mL) than other wasta streams. The second largest source term for accidents involving LWTS systems is considered to be the evaporator concentrates, which can reach 50 pCi/mL of Cs-137 under current design specifications. Another large source is the concentrator feed make-up tank overheads fed to LWTS at a concentration of 19 pCi/mL of Cs-137. Vasta streens other than these sources will be reduced by various decontamination factors associated with the LWTS processing equipment to levels significantly below 19 pCi/mL ... For the HEPA filter failure and the filter fire, the nuclida distribution is assumed to be represented by those nuclides found in the 8D-2 supernatant. The nuclidas end activities released for each accident are given in Table H.9.6-1 through H.9.6-7.

H.9.6.3 OFF-SITE RADIATION DOSES All seven accidents analyzed will reuult in the off-site release of radio-activity. In the case of spills, the material released would primarily be in aerosol or vapor form resulting from the evaporation of the aqueous phase. A HEPA filter fire would result in an off-site release as would the loss of the V0G HEPA filter integrity due to high dust loading or vetness. In the case of fire, the integrity of the downstream filter'is assumed to remain intact with a diminished DF of 100. For filter failure, the downstreen filter is not CAF0778:3RM .

,. DC:91:0055 ATTACHMFRP C May 29, 1991 WNS-SAR-005 Rsv. 2, Draft C l

I' The dose to the maximally exposed off-site individual is given in Table H.9.6-1. A spill of 4500 L of evaporator bottoms and a release of 8.3E 02 curies to the environment is calculated to contribute 21 mrem over the two hour exposure period. The dose conversion factors (DCFs) used for these calculations are taken from WDP (1987) .

H.9.6.5 EUlm OF THE INTS FIFE CHASE AND/0E THE FIFE CHASE IJAVING THE 01-14 BUIIRIBC As discussed in Section H.9.6.4, rupture of these lines can be caused by a variety of external sources. At the time of this pipe rupture. Tank 35104 is assumed to be at capacity and the evaporator feed pump (71-P-01) is actively pumping decontaminated supernatant to t.he evaporator at the rate of approximately 200 litres / min. The pump is again assumed to be operating unnoticed for one hour following the rupture rpilling 11,000 L of decontaminated supernatant into the STS trench. The trench is conservatively assumed to be open to the atmosphere at the time of the spill. The decantaminated supernatant is assuwad.to be relcased at-a temperature of 27 C with evaporation taking place at an average temperature of 20 C. The waste stream is assumed to have a total activity of 10.9 pCi/mL with the Cs-137 concentration at 3.21 pCi/mL .

As in section H.9.6.4, the probability of such a rupture is unlikely.

1 Emergency procedures exist which allow for systematic shutdown in the event of I an accident. In addition, the process line itself is routed through a sealed concrete trench. Therefore, releases to the atmosphere are severely restricted by the enclosed trench when compared to free atmosphere, as was r.ssumed in the off-site dose calculation. The dose to the maximally exposed individus1 is 0.86 mren (Table H.9.6-2). A calculation was also performed considerin5 the CEMUT overheads as the source. The off-site dose from this source was calculated at 0.32 area (Dooley, 1986).

See also section H.9.6.11 for the analysis of a spill of evaporator cleaning solution.

CAF0778:3RM .

, DC:91:0055 ATTAO NFNP C May 29,1991 UVNS-SAR-005 Rev. 2, Draft C 127 ft, elevation and is positioned over many of the lines feeding the evaporator. As stated above, damage or disruption of these lines is assumod to have little or no additional impact since most system valves can be activated automatically or manually in the case of an emergency.

The zeolite ion exchanger is located at the 136 ft, elevation of XC-3. A portion of this tank is directly over the Spent Zeolite Tank 71-D 007. At the time of failure, the zeolite ion exchange tank and the spent zeolite tank are assumed to be at maximum capacity of 4,000 litres each, and to contain 4 curies of Cs-137 each, based on the maximum anticipated resin loading of 1 pCi/mL. The zeolite ion exchange tank is assumed to fall down, striking the spent zeolite tank and causing both tanks to fall to the floor of XC-3 at elevation 100 ft. The entire contents of 8,000 litres are assumed to spill f

l onto the floor. The dose to the maximally exposed off-site individual is t

l estimated to be 4.3 E-06 arem as a result.

The total liquid volume which could be accommodated on the XC-3 floor l (assuming the sump-is inoperable) before liquid would overflev into the CAA is approximately 18,000 litres. Therefore more than twice the volume of liquid associated with the maximum spillage resulting from vessel support failure can be accommodated and contained within XC-3.

H.9.6.11 RUFTURE OF IRTS TRANSFER PIPE AND SFILL OF EVAPOEAM CUMUWG 7 SOLIrrios As described in Section H.6.1.1.10, the evaporator located inside XC3 may require periodic cleaning to remove accumulated scale. This operation will be conducted by recycling approximately 3600 L of borated nitric acid solution (up to 2 H) inside the evaporator at 100 C. No significant concentration of this solution will take' place. The acidic liquid will then be cooled in the concentrates cooler, and diluted with two evaporator flush volumes (3600 L each) for a total voluma of approximately 11,000 L. After transfer to the neutralization tank, the pH will be adjusted such that reprecipitation will ( g_

not occur. The resulting solution will be sampled and returned to Tank 8D-2.

~

CAF0778:3RM  !

DC:91:0055 ATTACHMENT C May 29, 1991 WNS -SAR-005 Rev. 2, Dreft C j The probability of a spill of this solution is very low due to the short duration of the operation relative to other LVTS processes. However, the spent solution is expected to contain a higher amount of actinides than normal LWTS waste streams (Cwynar, 1991). Prior to the first cleaning of the evaporator the mass of uranium accumulated in the scale was estimated to be less than 110 kg (1.7% U-235). The total mass of plutonium is estimated to be less than 460 g. As much as 40 Ci of mixed fission products may also be present in the scale. To be conservative, this activity is assumed to be entirely from Sr-90. Following dilution with two flush volumes, the radionuclide concentration in the cleaning solution is expected to be less than 120 pCi/mL.

The accident scenario is similar to the one described in Section H.9.6.5. la this case, however, pump 71-P-06 is assumed to be pumping the dilute cleaning solution from Tank 5D-15A2 to Tank 35104 when the pipe rupture occurs. The entire 11,000 L are assumed to spill to the transfer trench. Seventy-three liters of liquid are conservatively assumed to evaporate during a two-hour release period. A partition coefficient of.1000-between liquid and vapor phases was applied to the activity in the evaporated liquid.

The dose to the maximally exposed off-site individual, using dose factors from WDP-065 (Rev. 2), was extimated to be 120 mram (Table H.9.6-7) . The dose to the maximally exposed on-site individual (assumed to be 100 m from the spill) was estimated to 1. .t.2 rom.

I Non-radiological releases from this accident (nitric acid and NC )x would result in concentrations below the applicable TLV-TWA at the site boundary.

H.9.7 KINDEL ACCIDEllTS A number of minor accidents such as valve failures, pump failures, loss of control panel signals for various operating conditions and mixing of acid and caustic were analyzed. No single failure or series of multiple failures produced results comparable to the major accidents above in terms of dose to the maximally exposed off-site individual. Inadvertent mixing of acid (HNO )3 and caustic (NaOH) in significant quantity could create an explosive mixture CAF0778:3RM .

DC:91:0055 ATTAO NFRr C May 29, 1991 UVNS-SAR-005 Rev. 2, Draft C capable of producing severe injury and possibly death. Administrative procedures will adequately protect against such an occurrence. Also, the acid and caustic addition to Tanks 7D 2, SD158, 5D15Al and SrL5A2 is such that both positive displacement pumps cannot be running simultaneously due to a manual interlock.

R.9.3 PorarrIAI. POR NDSJtAR CRITICALITY Decontaminated supernatant fed from STS will be evaporated and thus concen-trated by LWTS processing. The resulting concentrations of U-233, U-235, Pu-239 and Pu-241 must be determined based on the dasign limitations of the evaporator's ability to concentrate decontaminated supernatant. The concentrations of fissile radionuclides at the processin5 limit of 46 w/o solids is shown in Table H.9.8-1.

Table H.9.8-2 shows the maximum allowable concentration of fissile material that is administrative 1y allowed in a process vessel. To ensure these limits are not exceeded within the confines of the Liquid Waste Treatrant Systar.,

batch samples will be collected and analyzed for fissile content at various points in the process. The total maximum concentration of all fissile material in evaporator concentrates is not expected to be greater than 6.6 E-04 grams per litre (O'Ahoofe, 1985). This is well below the most limiting case from Table H.9.8-2, where the uranium concentration is 0.05 g/L and the plutonium concentration is 1.63 g/L, for a total concentration of fissile isotopes of 1.68 g/L. All other combinations of fissile uranium and plutonium nuclides have suberitical concentration limits greater than this value.

The accumulation of fissile materials in evaporator sea.le is discussed in the next section.

CAF0778:3RM _ _ _ _ _ _ _ _ _ _ _ _

1

, DC:91:0055 ATTACHMPRP C May 29, 1991 WNS-SAR 005 Rev. 2, Drcft C H.9.8.1 EVAPORAT0E ACID WASH Based upon trend analysis of process sample data (Cwynar 1991) following supernatant treatment, it was conservatively estimated that not more than 460

-grams total of Pu 239 plus Pu-241, 1.88 kg of U-235 and 105.9 kg of U-238 could be present in the avaporstor. To ensure that no potential for criticality exists, bov' 'i v criticality calculations were performed by Yuan (1991) using the SCALE / KENO V computar code system and confirmsd by Caldwell j (1991) using the MCNP computer code. A right circular cylinder (radius 60 cm) I with sides and bottom of 1/2 inch stainless steel (i.e., stainless steel can) was used to approximate the reboiler portion of the evaporator for the mathematical staralation.

It was further assumed that all the fissile material was dissolved in water in j a hemisphere on the bottom surrounded by water. Yuan performed a search for the highest k,gg assuming a fixed mass (460 grams of Pu-239) and only varied 1 the radius of the hemisphere which varied the density of the Pu for a given total mass. Caldwell further studied the sensitivity of the k,gg calculations by varying the amount of fissile plutonium by i 104. These results established the stabiaity or relative sensitivity of the k,gg to changes in the fissile mass. Additionally Caldwell examined the effects on k,gg if the stainless steel can was reflected by water below the hemisphere in the can.

These calculations provide assurance of suberiticality evwn under the assumption of the avsporator cell flooding. Similar calculations were performed using both Pu and U.

Results of these calculations show that the k,gg + 2a of the evaporator containing the fissile material remains less than 0.95.

l t

l CAF0778:3RM DC:91:0055 ATTAOMENT C May 29, 1991 WVNS-SAR-005 l

Rav. 2 Draft C l

Realistic considerations of mathematically characterizing the results of acid washing / cleaning the evaporator would include estimates of additional salts being solubilized as well as a much more dilute solution resulting in further reductions of k,gg. Both Yu.an and Caldwell performed calculetions considering the total amounts of fissile material dissolved in water in the bottom of the evaporator (reduced right circular cylinder of radius 60 cm and associated lO height to optimize density and neutron leakage). The results confirmed the negative effects (lowering the k,gg) of the dilution and changing from the j hemispherical geometry to the cylindrical geometry. Parks and Dyer (1991) provided an independent validation of the analytical approach.

The conservative mathematical assumptions used for the bounding criticality calculations envelope the maintenance activity of acid washing / cleaning the evaporator.

CAF0778:3RM - 81

vu m uv.,,, m mm u ,4  % wy , -

WNS SAR-005 R3v. 2, Dreft C REFummarTt POR SECTION H.9.0 American National Standards Institute, N46.1, 1980, Guidance for Defining Safety Related Features of Huclear Fuel Cycle Facilities.

Caldwell, J. T. ,1991, Letter ZW:91:0031, "K,gg Calculations for WVNS",

dated March 11, 1991.

Cwynar, J. C. ,1991, latter WD:91:0324 (DC:91:0027) to T. J. Rowland,

" Conclusion on Fissile Material Accumulation fu the IRTS Evaporacor,"

dated March 22, 1991.

DRAVO Engineers, Inc., 19ES, Existing Wasta Evaporation Performance Evaluation Repcet on Performance Study of Evaporator.

Dooley, D. A., 1986, Calculation Set 86 002, LWTS SAR Accident Analysis, n_ der, J. C. et.. al,1986, "A Guide to Rs.diological Accident Considerations for Siting and Design of DOE Nonreactor Nuclear Facilities," Los Alamos National Laboratory, IA-10294-MS , UC-41.

LWTS Process and Instrumentation Diagrams, 1985, Drawings No. 901-D-020S2 through 020S4, 901-D-021 through 026 and 901-D-059, Ebasco Services Incorporated.

Marchetti, S., 1987, Memo to Distribution, " Anchor Bolt and Welding Problems,"

-Memo AD:87:0099.

Nuclear Fuel Services, Inc., 1962, Final Safety Analysis Report, Volume I.

O'Ahoofe, K. A. ,1985, Memo to J. C. Cwynar, dated March 26, 1985, " CSS Criticality Safety Evaluation," FB:85:0072.

Parks, C. V. , and Dyer, H. R. ,1991, Letter Report to P. J. Valenti, " Visit to WVDP to Discuss Criticality Safety Concerns Associated with Fissile Material Accumulation in the-Integrated Radwasta Treatment System Evaporator. February 13-14, 1991* da':ed March 19, 1991.

Peterson, J. M. ,1986, Letter HE:86:0025, J. M. Peterson to D. K. Ploetz, dated February 4, 1986, " Correlation of Exposure Rate with Radionuclide Inventory on a Loaded HEPA Filter."

Process Description and Central Concepts (LWTS) Revision B, Ebasco Services Incorporated, 1985.

US Nuclear Regulatory Commission (NRC), 1982, Regulatory Guida 1.145,

" Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Revision 1.

CAF0778:3RM

,. DC:91:0055- ATTAQlMENT C May 29, 1991 WVNS SAR 005 R:v. 7, Draft C West Valley Nuclear Services Co. Document Number 025, Design Criteria for Liquid Waste Treatment System, Rev. O,-8/85.

West Valley Demonstration Project, 1985 Vol'eme I, Project Safety Analysis Report, Project 0verview and General Information.

West Valley Demonstration Project, 1985a, Volume II, Project Safety Analysis Report, Existing Plant and Operations.

West Valley Demonstration Project,1987, Radiological Parameters for Assessment of West Valley Demonstration Project Activities," WVDP.065.

Yuan, Y. C., 1985,14tter HE:85:0156, Y.'C. Yuan to C. J. Robert, dated July 31, 1985, " Evaporation from a Liquid Spill."

Yuan, Y., 1991, Memo FB:91:0081 to C. J. Roberts, " Criticality Evaluation IRTS Evaporator Draft, Final Report," dated April 11, 1991.  ;

I 1

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CAF0778:3RM

.. :91-0055 MTACI ENT C by 29,1991 WNS-SAR-005 Rov. 2, Drcft C TABI2 H.9.6-7  !

t DOSE TO THE MAXIMALLY KIFOSED OFF-SITE INDIVIDUAL FOR THE SIGNIFICANT l EADIONUCL1 DES EVAF0 BATED FICEE A 3 FILL OF EVAPORATOR CLEANING SOIRTION i FOIJMING EUFttRW OF A TEAM 5FER FIFE Activity Source Nuclide inSpi}1 Partition , Term C.E.D.g.

(C1) Coefficient" (C1) (area)

$r 90 4.0 E+01 1000 2.7 E-04 8.3 E 02 Pu-238 9.3 E+01 1000 6.3 E GA l_ 6.8 E+01 Pu-239 2.2 E+01 1000 1.5 E 04 Pu-240 1.8 E+01 1.7 E+01 1000 1.1 E 04 1.3 E+01 Pu-241 1.1 E+03 1000 7.7 E-03 1.8 E+01 '

TOTAL 1.3 E+03 8.8 E 03 1.2 E+02 '

1 The significant radionuclides are identified by the product of the source term (C1) and dose conversion factors (res/C1) for ground-level releases. l total dose areThose radionuclides which contribute more than 0.1% of the included. i I

2 Per discussion in text, 11,000 L of dilute evaporator cleaning solution are spilled to the transfer trench.

3 The partition coefficient is the ratio of the activity concentration in the liquid to that in the vapor above the liquid pool.

• 4 The volume _of liquid that evaporates over a two hour release period is estimatsd to be 73 L.

5 NOTE: 1 Dose factors are from WDP 065, Rev. 2 (1990) . The CEDE to the maximally exposed on-site individual will be a factor of 18 higher. '

v i

CAF0778:3RM

, , . ,,,,,w--- ^

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, DC: 91:0055 ATTACIBENT C May 29, 1991 WNS - SAR-005 Rev. 2. Draft C 1

H.10.4 NDEMAL IRTS OPERATIONS H.10.4.1 INTS PROCEDURES IRTS will be operated using procedures prepared, reviewed, and approved per the requirements of the WNS Policy and Procedures Manual. Procedures will be written to cover all aspects of IRTS operation including but not Ilmited to:

Tank 35104 and Pump P-01 Evaporator 31017 Distillate Surge Tank Concentrate Cooler Dist.illate Pump Cleanin5 Zeolite Ion-Exchange Operation (003)

Filter Law TDS Feed Tank and Pump Filter Backwash Tank Organic and Zeolite Ion-Exchange Vess.els (001, 002)

IRC Tanks and Sample Tank (009)

Rosin Hopper and loading Media into 001, 002, and 003 Ion Exchangers Unioading Madis from 001, 002, and 003 Ion Exchange Vessels Unionding Solids from Storage Tanks 006, 007, and 008 Concentrate Hold Tanks 5D15A1 and 5D15A2 and Pumps P-05 and P-06 Lagoon 2 and Interceptors Lagoon 3 and Recycle Water System Chemical Addition Tanks 14D7 and 14D18 Atmospheric Release Evaporator CAF0778:3RM 94 I

DC Re0055 ATTAOIMENT C May 29, 1591 FA 91:0049

. RS 91:0024 ATTACHMENT B RESOLUTION OF ITEMB FROM REFERENCE 2 fATTACHMENT C)

Itae No. Resolution 1 Page 10; H.9.2 has been changed to H.9.6.

2 Page 29; Now reads: "Since these solids will cot dissolve in water alone it is necessary to use up to 2M nitric acid ($12%) to convert them into a soluble form."

3 Page 29; Now reads: "However, approximately one gram per liter of boron (as boric acid) may be added to act as a neutron poison for fissile isotopes which may be present."

4 Page 29; Now reads: "The solide increase radiation background  ;

near the evaporator, lower its boiling capacity and accumulate fiosile isotope in the evaporator scale or sludge ir the bottom head."

l 5 Page 30; The HEPA's are acid resistant and if the concentration i

was 66% higher, then the concentration of 8.3 ppb. (1.66 x 5) l would still be negligible. Past experience with the VOG system with much higher concentrations for longer periods of time (3 to 6 months) has produced no discernable HEPA filter degradation.

6 Page 30; The trigger levels for future evaporator clean out will be addressed in the next revision of the LWTS SAR prior to resumption of sludge wash processing.

7 Page 69; A new bounding accident has been analyzed and is discussed in Section H.9.6.ll, " Rupture of LWTS Transfer Pipe and Spill of Evaporator Cleaning Solution."

C Page 70; As rtated in Section H.9.6.9, "VOG REPA Filtcr Fire" the

" filter material is fire resistant." The accident as analyzed results in an estimated dose to the maximally exposed off-site individual of 3.4 nres. Past experience based upon filter changes shows that no significant source of_other material (lint) which could further support combination accumulates on the filter.

9 Page 77. This section (H.9.6.ll) has been replaced by a new accident analysis and thus the comment is no longer relevant.

10 Page 80: Additional criticality configurations were considered and are discussed on the next page of this revision (p. 81).

11 Page 81; The criticality calculations were conservatively modeled. Following dissolution of the material in the evaporator, subsequent leakage of the dilute solution would only further reduce the k,,,.

RLW4300 SEA-163 B-1

DC 3:00SS ATTAGP1ENT C bby 29, 1991 FA 91:0049 RS:91:0024 12 Page 81. As now stated in Section H.9.6.11 "After transfer to the neutralisation tank the pH will be adjusted sucF that reprecipitation will not occur. The resulting solucion will be sampled and returned to Tank 8D-2.*

13 Table H.4.1-2 footnote has been removed. There are no changes on this table for this revision (page has been removed).

14 Page 26, H.6.1.1 Changes han been removed. There are no changes in this section for this revision (page has been removed).

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4 # West Valley Project Office DW:91:0341 May 15, 1991  :

y f p.O. sox 191  !

West Valley, NY 14171 l

May 15, 1991 l Mr. J. J. Dtrygy Jr., President I and General Manger j West Valley Nuclear Services Co., Irc.  !

P. O. Bcx 191 J West Valley, New Yor!c 14171 i

ATIENTICH: J. T Knabenschuh, Vice President & Emrb.uu=4tal Safety, Health

& Quality Assurarca Manager {'

SUa7ET: I West Valley Project Office (WVB3) Ccments en Pavised IWrs

- (SAR-005) __.1 l cear Sir:

IMM belcw are the WVEO n . .--4ts en the subjec*: SAR. %ese cu.a.=4ts were l aie==d with members of your staff and are being transmittai to you as a retter cf racctd.

7~4am Mo.

_f. o Pg 10, H.4.4, ewvni paragraph; Reference to accident. analysis in H.9.2 - locks like accident analysis is in H.9.6.

, 2. o Pg 29; What is the m@alirx3 Molarity for 12% Nitric Acid?

3 o Pg 29; ne semerce ~% that the additien of Ecrun mad to prudent doesn't ccnnit to anythirq. Will Ecran be aMM? If so, how much is a little?

9 o Pg 29, first revised paragraph, secord sentence; ...ma - late l fissile isctcpes in the evaporatar scale er sludae in the het*a hMd.

[ o Pg 30, seccrd revised paragraph, last sentence; Are the HEPA's acid resistant? Will they be unaffectal if the u.u.=hations .gy,; 66%

hisber?,

I r-oq. -

What is the trigger level fer perfamirq arx:ther system flush? '

7 o P9 69, n.s.s.1; An-*nt (7) is an ac:hi Wash spill. What abcut an acid " diversion: to another rw pipiry system? Say thretx;h a

. failed valve allig M ? Pbrther, what dces the am-4 dant analysis

==lar-"4cri w a lock like? How is it da-484 that the spill is

' the caly credible an-idant a"^-4atai with the cleanirs su.=ss?

How ab:nt an evaporatcr overfill? Ecw ateut acid diversien to a zeclite bed?

,. /

C-1

DC:UL 0055 ATTAC10ENT C May 29, 1991 J. J. Buggy May 15, 1991

$7*/*4 /l*.

8 o Pg 70, H.9.6.3; Will a filter fire be enhanced with 1903 ^MM to the off-gas stream?

s9 o Pg 77, H.9.6.11; "No w a.=6L aticn..." is too limiting. "No sigrdficant u-Laticn...or "Csr.=ukaticn should he limited to..." veuld be preferred.

/O o Pt. 80, criticality configurations; Sbculd merttion be ::ade of all of the cmfiguraticos considered? A listing? With a notation that nana were mcre exoservations?

// o Pg 81; Is it en to censider criticality if the solutions leaked out of the evapcratcr to a sump?

/ "2 o Pg 81; Will Aerosipitatirvi of fissile material follcuixxg neutralintion cause the fissile material to get " hung up" in other parts of the system cn the way back to 8D-2? Has this been evaluated?

/ s3 o TAELE H.4-1 FwL.sste (1); Khat dces th.s i mean "...rct ir1 W ...to avoid double counting." Where are they included?

/yo H.6.1.1; I.HIS is divided into 9 r u w w ... Khy is this revisai unless something below was added?

Please contact me en extensicn 4780 if any of these -nts reg 2 ire additional clarificatien.

Sircerely, j'i &

T. J. , Wm West Valley Project Office cc: J. E. Solecki, DOE-ID D. B. Engelman, DOE-ID ,

CSL:018:91 - 0785:91:10 CEL/sl' -

C-2 '

A '

• WD:91:0596 v West Valley 'o 80 19' E:

r Nuclear Services C0mpany 10C0fD0 fated

$ nun,nn3m % nin ein

\

May 31, 1991 Mr. T. J. Row)and, Director

' West Valley Project Office U.S. Department of Energy MS-DOE P.O. Box 191 West Valley, Nnw York 14171-0191

Dear Mr Rowland:

Attention :<J..A. Yeazel:

SUBJECT : Response to NRC Cotaments on Cement Recipe REFERENCE : 1) DOE letter number WSK:015:91, T. J. Rowland (WVPO) to J. J. Buggy (WVNS), "NRC Response to i the West Valley Cement Recipe Data Transmittal",

{ dated May 7,-1991 ihe purpose of this letter is to provide WVNS's response-to the NRC comments (transmitted by WVPO in reference 1) regarding the

January 1991 WVNS Development Of A Cement Waste Form For Decontaminated Tank 8D-2-Sludge Wash " document. These responses are conta1ned in attachment "A" of this letter and are based on input from Analytical & Process Chemistry (D. J.-Fauth, L. E. ..

Michnik, R. A. Palmer), IRTS Process Control Engineering (J. C.

Cwynar) and J. Mahoney Vitrification Process Development.

This letter-satisfies commitment DW:4410. If you have any questions, please contact the undersign on extension 4382.

P.S. KlaT11an, Manager Vitrification Test Group West Valley Nuclear Services Co., Inc.

EM:91:0035 PSK:psk cc:

.C. Leek R. Provencher

.J.A. Yeazel EM:91:0035 u.-., o, Westmpouse Ele:tne Corporanon Ozy: 9/ /o

.- -- .. - . - - - ~ . . - - . - . . . - . .- - - - -- _

a ATTACEMENT A REVIEW COMMENTS ON JANUARY 1991 WVNS DEVELOPMENT OF A CEMENT WASTE FORM FOR DECONTAMINATED TANK 8D-2 SLUDGE WASH Part Is Clarification of " Technical Position on Waste Form" (TPWF)

Revision 1 NRC QUESTION :

TPWF 3ection C.2.a - Compressive Strength Requirements.

WVNS has proposed to use two-inch cubes made in accordance with ASTM C109 for the laboratory scale specimens instead of cylinders as  !

directed by ASTM C39. We have determined that this is an acceptable l alternative, because the waste forms will not have any particles large

-enough to dominate the compressive strength test results. The waste t

is a filtered liquid, concentrated by evaporation t, 33 wt% solids, and is unlikely to have any particles at all other than small crystals. If the cement is thoroughly mixed, the uncured waste form is expected to be a smooth paste without particles. Care must be taken not to exceed the load application rates specified.in Section 9.6.3 of test method ASTM C109.

WVNS RESPONSE :

Agree (with the following reservation). Because of radiological contamination-concerns, we are evaluating alternatives to cube sanding as specified in ASTM C109. NRC concurrence will be requested prior to any such change.

NRC QUESTION TPWF Section C.1.e - Leach Testing The plan and the Topical Report should include the projected concentrations of radionuclide tracers, and the basis for selecting these' concentrations.

WVNS RESPONSE.

l L Agree. This test is currently being designed. The information will be' supplied after the analytical chemistry method has been developed and approved. This method will be issued by June 14, 1991.

NRC QUESTION :

TPWF Appendix A - Cement Stabilization

-Section II.B - Compression The plan and the Topical ~ Report should specify the number of two-inch cubes that will be prepared for compressive strength testing. The i

i

.s TPWF recommends a minimum of 10 samples for statistical meaning.

The Plan and the Topical Report should include developing a correlation between the compressive strength data using ASTM C109 compressive strength tests and the penetrometer or other tests to be used in PCP verification testing.

WVNS RESPONSE :

Agree. The Test Request / Procedures, contained in the notebook (previously sent) specify the number of cubes reg'lired for each test.

Regarding penetrometer tests, the penetrometer data is being used only as an indicator to determine initial set times. Compressive strength tests are being performed on all specimens.

NRC GGESTION :

TPWF Appendix A - Cement Stabilization Section II.D. - Irradiation The TPWF recommends irradiation testing under two conditions:

- the specimens contain ion exchange resins or other organic media, or the expected cumulative dose is greater than 10E+9 rads.

The sludge wash waste forms will not contain ion exchange media, and the total organic content is expected to be less than 500 ppm.

Previous experience with the supernatant cement solidification indicated that oxalates at concentrations less than 39 ug/g appeared to retard the set (Reference 1). However, there is no indication that this concentration of organic would cause the waste form to be vulnerable to irradiation degradation. The sludge wash simulant will be prepared with almost 1% orga':c acids. Irradiation testing of the supernatant specimens indicated that they were not degraded (Reference 2). Based on this previous data and the preposed sludge wash simulant formulation, we have determined that irradiation testing of the sludge wash qualification specimens is not required.

WVNS RESPONSE :

Agree.

NRC QUESTION :

TPWF Appendix A - Cement Stabilization Section II.E. - Biodegradation Biodegradation testing of qualification specimens is recommended by the TPWF only if the waste forms contain carbonaceous materials. As stated above, the sludge wash qualifications specimens wil' be prepared with 0.5% organic constituents. Since the supernatant

i .. .

e specimens were spiked withfmore than this amount, and their-

,a biodegradation test results were acceptable, the Division will not require biodegradation testing of the supernatant specimens.  ;

WVM8 RESPONSE Agroe.

'Part III Plans for Evaluatina Sludce Wash Cement Recipe NRC CONNENT 1.- The plan should specify how many simulant cylinders will be prepared for correlation with the laboratory sized specimens.  :

WVN8 RESPONSE :

Agree. Per Test Procedure TP-030, Twenty-two cores will be obtained from.five full-size drums. These cores will be crushed and the compressive strength will be compared to laboratory sized specimens.

NRC CONNENT 2.- The plan should specify how may 1" x 3" cylinders will be prepared for the leach test. Although neither the TPWF nor ANS 16.1 specifies the number of specimens to be prepared for the leach test, it is suggested that test be performed on more than one specimen.

WVNS RESPONSE t' Agree. This information is included in Test Procedure TP-026. We plan to use three 1" x 3" cylinders for the leach test.

NRC QUESTION

3. The plan should specify how the confirmation cube will be prepared and whether it is from the sample collected in 1989 or the new composite sample collected in 1991. The plan should specify how the confirmation cube will'be cured. The plan should also provide the criterion used to determine whether the confirmation cube compressive strength is acceptable. We believe that this criterion should be that the confirmation cube strength is within the range of compressive-strengths of the simulated specimens.

WVN8 RESPONSE :

Agree. Confirmation cubes-will be prepared-based on a 129-inch heel using wash from pre-1991 sampled sludge and 129-inch heel with wash from 1991 sampled sludge wash. Criterion for-acceptable compressive strength will be established using data generated fe; the nominal waste simulant-in the multivariate tests performed to Test Procedure TP-028. Curing profiles for these cubes will be derived from data generated from a thermocouple instrumented drum in our 20-drum 4

_ _ _ _ , - + __ -

simulant run (TR-026).

NRC COMMENT

4. The plan should describe clearly how the concentration limits and operating limits testing will be done. A list of the recipe variations that will be tested, and how the limits will be salected, should be provided.

The curing regime should also be clarified. It is unclear how many cubes will be in a " set", and how long they will be oven- and ambient- .

cured before compression testing. It appears that the oven curing I temperature is 88'C; this should be equal to the centerline temperature for the supernatant full-size waste forns.

WVNS RESPONSE Agree. Our initial recipe variations and the number of cubes are i included in the Test Request TR-028 and Test Procedure TP-028. A thermocouple instrumented drum has been temperature profiled in accordance with SIP 91-01. The peak centerline temperature for this drum was 79'C. The centerline temperature returned to ambient temperature in a total of 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />. Our curing cycle for all specimens will include an initial 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> oven cure with the remainino curing taking place at ambient temperature. The Tent Requests /

Procedures, contained in the notebook (previously sent) specify the number or cubes required for each test.

NRC COMMENT Comparison of the Simulant and Cement Recipe The table shows a comparison of the supernatant and sludge wash simulants. Based on the comments addressing item 4 above, it appears that this formulacion could change slightly as a result of the concentration limits and operating limits testing.

The major differences between the supernatant simulant and the sludge wash simulant are mostly due to the reduction in solids content from 39% to 33%. Other differences, e.g., the decrease in nitrate and the increase in sulfate are explained by the sludge wash analytical data.

However, it is unclear why the organic acids, citric, oxalic, and tartaric, are reduced by a proportion greater than that of the solids.

Since these have been demonstrated to cause problems with the waste form set, their concentrations in the simulant should be selected to be conservative.

WVNB RESPONSE :

Agree. We havn looked over the recipe and are making minor adjustments. These will be included in the notebook material when finalized. Variations for organic constituents will be included and are included in the parameters of the multivariate testing (TP-0/8).

Variable ranges for the tests are being selected conservatively.'

2

West Valley nec. :- ser =tS TP 02s Demonstration Project " " " t o " ""*

• r Revision Date 05/2./91 Engineering Release //2098 4

c.

TEST FROCEDURE PROCEDURE FOR DEVELOPMENT OF THE NOMINAL RECIPE FOR l l

CEMENT SOLIDIFICATION OF SLUDCE WASil LIQUIDS PREPARED BY . 488// 6 /%._ M. N. Baker Cognizant Engineer APPROVED BY V D. C. Meess Cognizant Systern Design Mana er j .1 N APPROVED BY '7 4 ul. /.//teu 'p[i. D. L. Shugars

'/

QualityAsjutanceManager APPROVED BY E) My P/7,5 82/@D. J. }larward

,rdiation and Safety Manager APPROVED BY k 6. thm J. C. Cwynar Prdcese Condrol Engineering West Valley Nuclear Services Co., Inc.

P.O. Box 191 l UELOO49:3RM West Valley, NY 14171-0191

\W-1816, Rev.1 l .. .- . _ - .- . _ . - . . - _ _ _ , - . _ , . . ... -. - -.

WNS TP 025 Rev. O RECORD OF REVISION PROCEDURE If there are changes to the procedure, the revision nunber increases by one.

These changes are indicated in the lef t matgin of the body by an arrow (>) at the beginning of the paragraph that containt a change.

Example: *

> The arrow in the margin indicates a change.

Revision On Rev. No. Description of Changes Page(s) Dated 0 Original Issue All 05/24/91 W 1807, Rev. 1 i BEL 0049:3RM

. . - _ . - _ . ..- . - -.-.- ..-. . ~ . - _ .

WNS.TP 0?$

Rev. O RECORD OF REVISION (CONTINUATION SilEET) i Rev. No. Description of Changes Revisio(n)on Page s Dated ,

1 i

l l

l l

I I

L UV 1807 Rev. 1 11 '

LEL0049:3RM

WVNS TP 025 .

l i PROCEDURE FOR DEVELOPMENT OF Tile NOMINAL RECIPE FOR CEMENT SOLIblFICATION OF SLUDGE WASil LIQUIDS Rev. 0 1.0 SCOPE

'1.1 This-work is required to develop a stable waste form for cement )

. solidification of Sludge Wash liquids which exhibits the characteristico required by 10 CFR 61, Code of Federal Regulations Title 10. " Licensing Requirements for Land Disposal i

of Radioactive Waste." and the USNRC Branch Technical Position on l Waste Form, revision 1, dated January, 1991, 1.2 - The recipe for coment solidification of supernatant t (DOE /NE/44139-49) will be used as a starting point for this test ,

procedure.

1.3 Work will be performed with a simulant representing the actual 4 waste liquid to develop the " nominal" recipe for solidification of a sludge wash liquids.

1.4 A prerequisite for all work will be the decision by the IRTS Restart Task Force as to the actual level of supernatant liquid in High Level-Vaste Tank 8D 2: 129 inches. -

1,5 A prerequisite for all work will be the determination by the IRTS Restart Task Force of the expected amount of Sulfate in the Sludge Wash Liquid. The composir. ion of the " nominal" recipe, based on letter No. EKs91:0047, is given in table 1.

1.6 Work will be performed usin5 cubes 2" x 2" x 2" cast from'a  !

simulant / cement-mixture produced in the Analytical Chemistry Lab, BELOO49:3RM 1-

Wl45-TP 02S i

• Rev. 0 1.7 The " nominal" percent olids (by weight) in the waste 11guld will be determined.

1.8 Tha " nominal" range of Calcium Nitrate recipe enhancer to be blended with Portland Type I cement will be det erinined.

1.9 The " nominal" water to cement ratio will be calct'ated as follows:

I "Ib' I """ *)

• 02 "" " II"C' "}

W/C - -(1 Calcium Nitrate fraction) x ( Weight of cement blend) 1.10 Determine the " nominal" amount of Antifoarn recipe enhancer to be added to the liquid mixture.

1.11 Determine the " nominal" amount of Sodium Silicate recipe enhancer to be added to the waste /coment mixture.

1.12 The maximum ,;ractical compresnive strength of the waste form will he determined.

1.13 The effects of variable recipe parameters on the " gel time" and free liquid of the waste mixture will be evaluated.

1.14 The " nominal" recipe developed herein will be scaled up and qualified in accordance with WNS TRQ 026. It will also serve as the "non.inal" recipe for Process Control Plan parametric window tests being performed under WNS TRQ 028, 2.0 DEFINITIONS AND ABBREVIATIONS 2.1 Definitions Cement - Dry Portlan:1 Type I cement in accordance with ASTM Standard C 150-85.

BELOO49:3PJi ___

V VVHS TP 025 Hov. O Cement Blend - A homogenous mixture of Portland Type I cement with -)

a percentage of technical grade flake or granular form calcium nitrate with NO ammonium nitrate.

Cast - A specimen mixed in a poly bottle and then poured into a mold.

Cube A 2" x 2" x 2" cast specimen.

2.2 Abbreviations ACM - Analytical Chemistry Method A&PC - Analytical & Process Chemistry ACP Analytical Chemistry Procedure ,

CSS - Cement Solidification System IRTS Integrated Radwaste Treatment System IUP Industrial Work Permit PCE - Process Control Engineering QA - Quality Assurance R/S - Radiation & Safety TDS Total Dissolved Solids 3.0 RESPONSIBILITIES 3.1 - Analytical & Process Chemistry performs all work in this Test Procedure.

3.2 Process Control Engineering (PCE) provides technical direction, and compares the test data to the Test Request.

l 3.3 Quality Assurance provides surveillance to ensure that the requirements of this test procedure are satisfied, and verifies that portions of the test (where independent -verification is required) were performed.

BELOO49:3RM g

WVNS TP 025 Rev. 0 3.4 Radiation & Safety monitors radiation and contarnination Icvels 4.0 TOOLS, EQUIPMENT, COMPONENTS, AND REFERENCES 4.1 Tools and Equipment

- "LIGilTNIN" Model TS 1515 lab mixer with high shear impeller

- 2" x 2" x 2" plastic cube molds 1

500 mL-poly bottles

- 250 mL poly bottles

- 20 mL scintillation vial

- magnetic stir. plate and stirring bar l

timer top loading analytical balance

- 'Forney Model PT-40 DR compressive strength testin, machine with-h'-draulic power unit and capping set

-4.2 Rear,ents 2

- Portland Type I cement per ASTM C 150 85

- General Electric AF9020 antifoam emulsion

- Sodium Silicate solution: Water-based solution with 28.5 to 29.5 percent SiO2

-- Powdered calcium nitrate BEL 0049:3RM 4-V

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WNS TP 025 Rev. 0 4.3 Components

- Despatch Series 16000 Envitonmental Chamber fully operational 4.4 Re fe renc e s 4.4.1 EP 11 001. Test Control 4,4.2 EP 11 003, Development Test Control 4.4.3 WNS TPL 70 011, Test Plan for Waste Form for Cement Solidification of Sludge Wash Liquids

)

4.4.4 WNS TRQ 025, Test Request for Development of the Nominal  :

Recipe for Cement Solidification of Sludge l' ash Liquida 4.4.5 ACM CEMPREP 4801, Preparation of Cement Samples in the Radiochemistry Lab wr1tten by C. W. McVay, et. al. '

4.4.6 ACP 7.2, Safety Practices for the Analytical & Process Chemistry Department ,

4.4.7 WDP 010, WNS Radiological Controls Manual 4.4.8 WDP 011, WNS Industrial liygiene & Safety Manus 1 4.4.9 USNRC Branch Technical Position ~on Waste Form, revision 1, draft dated December 1990 4'.4.10 ~ ASTM Standard C 109, Standard Test Method for Compressive Strength of flydraulic Cement Mortars (Using 2-inch or

• 50 mm cube specimens) l  !

' BEL 0049:3RM . , - . , -.-_....- -.-.-... . - ,- .- - - . - - - ---._-_ - .-_.......-_-._.-. - . . . . _ . . -

WNS TP 025 Rev 0 4.4.11 ASTM Standard C-617, Practice for Capping Cylindrical Concrete Specimens 4.4.12 ASTM Standard C-470, Specification for Molds for Forming Concrete Test Cylinders Vertically 4.4.13 ASTM Standard C 150, Specification for Portland Type I Cement i

5.0 CENERAL INFORMATION l

5.1 Performance of the " nominal" waste form developed under this procedure will be qualified under Test Request WNS-TRQ-026, and

-Test Procedure WNS TP 026, 5.2 -Quality Assurance should be notified prior to commencement of activities; in order to perform surveillance (s).

5.3 OPERATORS Sil0ULD_ PERFORM FREQUENT CliECKS ON SYSTEMS THAT ARE 1

TURNED ON OR SHUT DOWN TO ASSURE TilAT THE SYSTEM DOES WHAT IS EXPECTED, I.E., WATER FLOWS, PRESSURE RISES, ETC. IF THE REQUIRED ACTION THAT IS SUPPOSED TO HAPPEN DOES NOT HAPPEN, (1) STOP - DO NOT PERFORM THE NEXT STEP, (2) SECURE THE SYSTEM IN A SAFE MODE, AND (3) NOTIFY THE COCNIZANT A&PC SCIENTIST OR COCNIZANT ENCINEER IMMEDIATELY, ,

6.0 EMERCENCY RESPONSE

'6.1 For emergencies in the A&PC Lab, responses will be as' directed by ACP 7,2 and 1VDP 011.

6. 2' For emergencies elsewhere in the plant, responses will be as directed by WDP 010 and WDP 011, 3EL0049:3RM 6-

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_ _ _ _ _ _ _ _ ~ . _ _ _

WVNS TP 025 Rev. 0 7.0 DETERMINATION OF THE " NOMINAL" PERCENT SOLIDS IN Tile WASTE

-This determination will be made by the IRTS Startup Task Force prior to baginning this work.

8.0 DF. TERMINATION OF Tile " NOMINAL" EANCE OF CALCIUM NITRATE RECIPE EUllANCER TO BE BLENDED WITil PORTIAND TYPE I CEMENT 8.1 starting with the original recipe for encapsulation of Decontaminated Supernatant, test the perfcrmance of the waste form at varying percentagts of Calcium Nitrate in the cement blend.

8.1.1 Prepare a cube using the " nominal" blend ratio of 5.7 percent Calcium Nitrate in accordance with ACM CEMPREP-4801.

8 .1_ . 2 Cure the cube at 8815 degrees celsius for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

8.1.3 After curing, remove the cube mold. '

8.1.4 Sand two (2) opposite cube faces until flat 8.1.5' Place the cube in the hydraulic press, and measure the pressure at the cube yield point. Record the pressure, and perform the compressive strength calculation per ASTM Standard C 109. Record on form VV 2301.

8.2 Increase tha Calcium Nitrate percentage to 6 percent,'7 percent, 8 percent, 9 percent, etc., up to 12 percent Calcium Nitrate in the cement blend.

8.2.1 Prepare cubes at each new cement blend ratio in accordance with ACM CEMPREP-4801.

I DEL 0049:3RM ,. . . a. _ . _ . . . - .. .- -_._.... _ _ __.. _ _ _ . . _ _ . _ _ -- .,-._ _ _ _ . . _

WVNS TP 025 Rev. 0 8.2.2 Cure the cubes, remove the molds, cap and perform compressive strength testing in accordance with section 8.1.2 through 8.1.5 above. j l

l 8.2.3 Record the gel time, penetration resistance,-and slurry density on form WV 2301, 8.2.4 Record t he presence or absence of bleed water on form WV-2301. If present, determine the pH.

i 8.3' Decrease the percentage of Calcium Nitrate in the cement to 4 '

percent.

8.3.1_ Prepare the cubes at this cement blend ratio in accordance with ACM CEMPREP 4801.

8.3.2 Cure the cubes, remove the molds. sand two opposite cube faces until flat and perform compressive strength testing in accordance with sections 8.1.2 through 8.1.5 above.

8.3.3 Record the gel tima, penetration resistance, and slurry density on form WV 2301.

8.3.4 3ecord the presence or absence of bleed water on-form WV 2301. If present, determine the pH.

9.0 CAICUl ATION OF THE NOMINAL VATER-TO-CEMENT RATIO 9.1 After the Calcium Nitrate fraction of the cement blend is determined (section 8.0 above), calculate the nominal water-to-cement-ratio as follows:

("* ' * "****} *( * ** 0* E'***'*"}

W/C - (1 Calcium Nitrate fraction) x (weight of cement blend)

BELOO49:3RM .

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_ ..,n__.,,,,,.m. ,. ., -.%..---.;.-,-,- - , .

E VVNS TP 025 Rev. 0 l

l 10.0 _ DETERMINATION OF THE NOMINAL AMOUNT OF ANTIFOAM 10.1 After the nominal Calcium Nitrate fraction in the cement blend and nominal water to ccment ratio have been determined, the nominal amount of antifoam-in the recipe is to be verified.

10.2 With a11'other recipo parameters remaining the same, or as l previously determined in section 8.0 and 9.0 above, prepare a cube I l-in accordance with ACM CEMPREP 4801, 10.3 Cure the cube at 88 15 degrees celsius for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

10.4 Af ter curing, remove the cube mold.

10.5 Sand two opposite faces of the cubo until flat.

10.6 Place the capped cube in the hydraulic press, and measure the t

pressure at the cube yield point. Record the pressure, and l perform the compressive strength calculation in accordance with ASTM Standart C 109. Record on form WV-2301.

11.0 DETERMINATION OF THE NOMINAL WEICHT OF SODIUM SILICATE RECIPE ENHANCER TO BE ADDED i 11.1 Af ter the Calcium Nitrate fraction in the cement, water-to-cement ratio, and amount of antifoam in the nominal recipe have been

. determined, the nominal amount of Sodium Silicate additive is to l' be verified.

11.2 With all other recipe parameters remaining _the same, or as previously' determined in sections 8.0, 9.0, and 10.0, p epare a cube in accordance with ACM CEMPREP-4801.

11.3 Cure the cube at 8815 do5rees celsius for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

BEL 0049:3RM 9-1.;-=...-__.-

- = - . - - - -

- - . .- - - - - - . - - - - - - - = - -

WVNS-TP 025 Rey, O l

l 11.4 After curing, remove the cube mold, j 11.5 Sand two opposite faces of the cube until flat.

11.6 Place the capped cube in the hydraulic press, and measure the p6 essure at the cube yield point. Record the pressure, and perform the compressive strength cniculation per ASTM Standard C.

109. Record on form WV 2301.

12.0 DETERMINATION CF Tilt MAXIMUM COMPRESSIVE STRENGTH 12.1 After the Calcium Nitrate fraction in the cement blend, amount of l Antifoam additive, amount of Sodium Silicate, water to cement l ratio for the nominal recipe have been verified, determine the maximum practical compressive strength of the vaste form.

12.2 A mean compressive strength in excess of 500 psi after 28 days curing is desired, as discussed in the Branch Technical Position, appendix A.IT.B.

12.3 Sufficient samples shall be prepared to determine the mean compressive strength as well as the standard deviation. A minimum of 10 samples shall be evaluated.

12.4 the compressive strength vs. time will be determined as discussed in WVNS-TP 026, section 7.0.

4 13.0 CURING 13.1 A curing temperature of 88 15 degrees celsius as required for Cement Solidification of Decontaminated Supernatant will be used

~for initial testing, 13.2 Then processing full-scale drums under Work Order 9100084. a drum was equipped with thermocouples and a temperature recorder. The BELOO49:3RM -

10 -

l' v - - . -r gg, ' + -

. . , -wwe, ,-y,_,,, ,[ .% 3ww,.y.u,_.<r,.w., - . .,y-.,-....,n., -

VVHS TP 025 Rev. O drum temperature as a function of time was plotted. The effect of )

curing at this temperature profile will be evaluated as discussed in the Branch Technical l'osition, appendix A.III.B.

i 13.3 For this procedure, the samples will be bagged and cured in an oven or temperature controlled chamber.

13.4' The chamber will be equipped with a calibrated thermometer and temperature readings will be continuously recorded.

13.5 All samples will be kept in scaled containers and/or poly bags during curing and storage, as discussed in the Branch Technical Position, appendix A.III.C. This is intended to simulate the )

environm6nt in a sealed drum.

1 14.0 DETERMINATION OF " GEL TIMES" 14.1 For all sampics cast in the A&PC Lab, the cube molds will be filled in accordance with ACM CEMPREp.4801, with a 20 mL saintillation vial filled for each eube. ,

f 14.2 Visually check for gelation of the cement / waste product in the

- scintillation vial,

a. Check for gelation every 5 minutes, and do not disturb the vial between these time intervals.

b. Celation is a subjective determination; however, gelled product is indicated when the 20 mL scintillation vial can be tipped slowly to a 90 degree orientation, and the cement product will not deform-or-flow, and will retain a line perpendicular to the horizon. ,

c. Bleedvater is NOT to be interpreted as a sign of incomplete gelation. Estimate the quantity and determine-the pH if not reabsorbed after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

BEL 0049:3RM -. . . _ ._ _.-_. _ . . _ _ _ . . _ - . . . . . . . _. _ . . . _ . . . - . _ _ _ . - _ . . - _ _ _ _ _ . . _ . _ _ _ _ _ _ . _ . . _ . _ . _ . .

TABLE 1 HVHS-TP-025 Rev. O REYMED SALT CONCHIRATIONS FOR THE " NOMINAL" SlMU1 ANT RECIPE BASED OH 128.5" liLEL C.QH5IIII)ENT FORHt)LA _ WEIGHT Sodium Nitrate Nat10 3 286 lbs.

Sodium Nitrite nan 0 2 272 lbs.

Sodium Sulfate Na2SO4 170 lbs.

Sodium Bicarbonate NaHCO .

3 Potassium Nitrate KN03 17.9 lbs.

Sodium Carbonate Na2003 48.4 lbs. ,

Sodium Hydroxide Na0H **10.4 lbs.

Sodium Dichromate, Dihydrate Na2Cr207 2H 2O 1590 g Sodium Chloride Nacl 1310 g Sodium Phosphate, Dibasic Na2HPO4 950 g Sodium Molybdate, Dihydrate Na2M o04 .2H20 226 g Sodium Tetraborate, Decahydrate Na280 4 7 10H20 122 g l

Citric Acid, Anhydrous CH0687 165 g Oxalic Acid, Anhydrous CH0224 129 g I Tartaric Acid, Anhydrous CH0466 180 g Water HO 2 H68.0 lbs.

TOTAL WEIGHT 2483.1 lbs.

Weight of Solids 815.1 lbs.

Waight Percent Solids 32.83 percent Note that Sodium Bicarbonate does NOT appear as NaHC03 at elevated pH's.

• • The Sodium Hydroxide (NaOH) value is an approximation to arrive at a pH of 12.1. This value may vary.

SRC4094/1

-,m.g 1 , . , , , . . . , , .m ,__ . . .

WD:91:0725

$st Valley a r. m / *1 7'd *

. Nuclear Semces Company ^"' " " * <" 'd" e incolD0tated MS M July 5, 1991 l I

l l

Mr. T. J. Rowland, Director West Valley Project Office U. S. Department of Energy )

1 MS DOE P. O. Box 191 West Valley, New York 14171 0191

Dear Mr. Rowland:

Attention: W"*T "K6tols-SUBJ ECT: Meeting, Minutes for Cement Waste Form Discussion, NRC Site Visit of June 6, 1991.

During the subject meeting, an agreement was reached for VVNS to record meeting minutes and forward a copy to the NRC through the WVPO. Attached are the preliminary meeting minutes. The minutes are termed preliminary until concurrence is achieved with the NRC. Please forward a copy of the minutes to the NRC for review and concurrence.

Very truly yours,

/

- M J. 14/Mahoney \

Acting IRTS Enginee ing Process Engineer West Valley Nuclear Services Co., Inc. ,

~

D. C. Meess, Manager IRTS Engineering West Valley Nuclear Services Co., Inc, CJ:91:0067 JLM: src

Attachment:

A) Preliminary Meeting Minutes for Cement Waste Form Discussion NRC/WVNS Meeting June 6, 1991 SRC4168 A s.u oa o' ,. Gj e, g2

... m.:,u m.o..m , os

. _ _ _ . . __._ _ _ _ _ .___m_-._ _ _ _ _ _ _ _ _ _ _ . _ . _ _ . _ . . _

+

ATTACHMENT A Eg,ELIMINARY MEETING MINUTES CEMENT VASTE FOPJL l DISCUSSION NRC NVNS MEETING JUNE 6. 1991

- Discussed the WVNS thought to have the ASTM committee address the use of j

. templates when performing cube compression testing. The current method only addresses sanding the surfaces as an accepted handling opcion.

- Acknowledged the update on curing conditions from the instrumented drum produced in the 20 drum run: 79 'C for a total of 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />.

Indicated the radionuclides of interest that will be specifically tested during the leachant selection test will be Cs 137, Sr-90, Tc.99, and Pu-238/239/240/241. The WVNS plan is to use actual decontaininated supernate left from the operation of STS and to add sufficient quantities of sodium j sulfate and sodium hydroxide to simulate sludge wash solution.

- Confirmed the minimum number of samples, per 5.10 of TPL 70-11 was 3 with '

the_ exception of 10 for compressive strength.

i

- Addressed the use of a penetrometer in qualification testing and the development of a correlation to key cement properties. WVNS emphaaized that all qualification work was based on compressive strength testing of .

cured prodect so no correlation to penetrometer readings is planned. The IRTS Orerations staff will probably use a penetrometer reading as an internat guide to the processability and acceptance of individual batches.  ;

- WVNS affirmed the word " verify" in section 1.4 of SIP 91 01 should  !

actually be " develop" as it pertains to the relationship between cube and core compressive strengths. .

Table 6.2.1 of SIP 91 01 was indicated as bein6 out of date. Test Exception (TE) TE SIP 91-01-06 corrected the names, formulas, and level of minor species (including organics) in the simulant. This led to a ger.eral -

discussion of the issue of forwarding TE's to the NRC. It was agreed that the revised base document, which will. fold.all ouratandi.ig TE's into the updated document, will be forwarded to Steve Moto'a, WVPC This will be accompanied by a one page write up that deta..a sno_unne 4ying causes that spawned the TE's.

- Selection of the parameters for analysis, as noted in 6.2.3 of SIP 9101 is-not specific. Mary Adams felt that some input _from the NRC could be forwarded in this area.- WVNS indicated that this work is essentially complete and that any alteracions or additions cannot be completed.

I l

SRC4168 b

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ATTACHMENT A i

- Although table 1 of TRQ 025 does not reflect the updated simulant composition (to agree with the update to SIP 91 01), it is not planned to update the table.- This reflects the fact that this preliminary recipe l development wot4 is complete, also that the test conclusion that the nominal recipe (from supernatant processing) is acceptable for continued evaluation would not change if the experiments were repeated at a slightly.

different simulant recipe.

- Mary Adams, NRC, wondered about the relationship between the VVNS Analytical-Chemistry Methods (ACM's) and the ASTM and ANSI test methods.

VVNS-explained that the ACM's are explicit instructions to the laboratory technicians on the steps to execute the reference tests (ASTM or ANSI) in the VVNS lab.

- It was agreed that the 4.3.15 of TP 026 would be updated to reflect tho correct test request.

- Sections 7.0 through 7.3, and section 14.0 of TP 026 will be reviewed by VvNS to verify the cylinder and cube totals specified match the values presented in table 2 of that same procedure.

- A pre release copy of fest Procedure WVNS TP 025 was forwarded to Mary Adams at the meeting, along with a pre release copy of the VVNS responses to the questions raised in the 4/22/91 letter from the NRC to WVPO. Each j individual response was reviewed. The use of a penotrometer was raised again. -WVNS will-clarify its use in the overall cement qualification and production process. The test request identifitd as 038 in the response letter will be updated to 028. 7

- Discussed some of the long term testing of supernatant cement that VVNS has begun. Mary Adams clarified the definition of "certain" waste forms (high organics, and/or ion-exchange beads) that ce quire long term surveillance.

- Tentatively agreed that the NRC TER can be performed from the Test Summary Reports (TSR's) that WVHS is scheduled to generate from the numerous-tests. Also felt that a draft copy of the structure of the updated CSS PCP would be helpful to the NRC, i

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unil es-I-11 Water L Solvent vs. Time _

-- casing top

~

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06/01/83 10/19/89 03/08/90 07/26/90 12/13/90 05/02/91 Cate

Well 85-I-9 Water & Solvent vs. Time

- kerosene

-- casing top

-__-I i l l l 3 - -

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I 06/01/89 10/19/o9 03/08/5. 07/26/90 12/13/90 05/02/91 l

Date I

l

d Well 06-I-1 Water & Solvent vs. Time

- kerosene

-- casitsg top

~T I I I I wa t y r- loveL I

1.9 -- - - - - * - - - - - - - - - - - - - - - - - - - - - - - - - - + - - - - - - - - ' - - - - - - - - + - -

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06/01/89 10/19/89 03/08/90 07/26/90 12/13/90 05/02/91 Date I _ _ _ _ _ _ _ .

i i

W.ll 80-I-3 Wat.r Level vs. Time _

'asing top w.t.r i.v.

3.7 - -

• 48-9 ,

w

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_L I I I i 1 06/01/09 10/19/89 03/00/90 07/20/90 12/13/90 06/02/91 Date

I Well es-I-4 Water Level um. Time -

casing top I l l I l water lol ve. L 4 - , . ~ .--

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l l  ! l l 06/01,99 10/19/89 03/0L/90 07/26/90 12/13/90 05/02/91 Data I

w_ _ - ._ . . .

. . . . _ _ _ _ . . _ _ _ . - , . . . - . . . ~

Well es-I-5 Water Level vs. Time _

- - - casing top I I l l l water let I (

3.9 -

2.9 - ' ---

1.9 -

0.9 -

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1 I I I i  !

06'01/09 10/1S/69 03/08/90 07/26/90 12/13/90 05/02/91 Date

i I

.1 i

1 Well eb-05-N Water & Solvent vs. Time

- kerosene 1

casing top j

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i l i water love.L 1

4 -- - < -

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06/01/89 10/19/09 03/00/90 07/26/90 12/13/90 05/02/91 Cate

Wells 89-13-W & 89-13-E -

09-13-W Water Levels vs. Time

--- casing top 89-13-E 3 - - - +-

• m+-ad.>

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_l l  ! l l l 06/01 eor 10/19/89 03/08/90 07/26/90 12/13/90 05/02/91 Date

. . . . ._ _ . - , .- - . _ _ . >_,.m---____ .._.m > _ . . _ _ . - . - . . . _ _ . -_ _ _ ._ _ .-.. . . - _ _ - - . _ _ _ . .

i l

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i i

Well 86-I-6 Water Level vs. Time -

coming top i I I I I I w.ter lov. ,

3.9 -- ' -* --

6 m

m 6-- -

m T $ w 4f h 4 & 6 gp 4 Ga M MW @ W@ WDM W 4 M W O %f @ n M 4 % @W 4* mMM SWWW W q W g 6 mmQ @ @ & M 4 m

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. . - - . . ~_. .. - , . .~- - . .- - - ...~.. =

Welle 89-14-N & 09-14-E Water Levolm -

14-H solvent and Solvent Level vs. Time casing top I I I l  ! I 14-N water 4 - - ' -"' '

- -', --J 14-E water -

M

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l Date

Welle 69-27-ti L 69-27-W Water anu ti univent Solvent Levels vs. Time

" - - casing top 1 I I I I 27-t8 we er b

o ---

27-W water 2 -

<'------.a---.-----.a---.e-------------.1 4. - - . . -- --- -- -

em . ~

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Well 09-27-H Water & solvent vs. Time

- korosene

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l l

l

l l Well 89 29-N Water & Solvent vs. Time - kerosene l

-- casing top l l l l l l water Itve L 3 - _

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A-

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i Well 39-29-E Water & Solvent vs. Time

- kerosene

--- casing top I I I I I i water leve

---

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l Date i

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Date

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Well 09-20-W Water & Solvent vs. Time

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October 1990 Average Stream Flow and Daily Precipitation 2

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3'v i- i i 4 i i i IM i i,,i 0 01 -Oct-9 0 0 8-Oct-9 0 15-Oct-90 2 2 -Oct-9 0 29 -Oct-90 Date Erdmann Brook 006 Outfall -

Franks / Quarry C M Precipitation

September 1990 Average Stream Flow and Dailf Precipitation 2

20 t

- 1.8 g - 1.6 15-  !

- 1.4 f ' C

.9_

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, , , , j i 3  ; i , , , 3 i 0 O , ; , , , , ,, , , , , 3 i 29 -Sep-9 0 0 8 -Sep-9 0 15 -Sep-9 0 2 2 -Sep-9 0 01 -Sep-9 0 Dole Erdmann Brook 006 Outfall Franks / Quarry C E Precipitation

.I August 1990 Average Stream Flow and Daily Precipitation 2

15 ,

1.1 -

- 1.8 J

13 -

] - 1.6 12 -

11--

l - 1.4 "g. C 10 1.2 94 y) b .

-h 5g '

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c 5 -0.6

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-- - i ') I 0.4 h \ i g 1 0 ._ s ^ _ .

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/, , , i i i 3 i 0 0 8 - Aug-9 0 15-Aug-9 0 2 2 -Aug-9 0 2 9- Aug-9 0 01 - Aug-9 0 Date Erdmann Brook 006 Outfall Franks / Quarry C E Precipitation

July 1990 Average Stream Flow and Daily Precipitation 2

12 ,

11 - 3 - 1.8 8

10 - f n

- 1.6 r

G l\

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June 1990 Average Stream Flow and Daily Precipitation 2

10 .

9_ /

/\ A\

_ .- 3.8

, j N. 1.6 8--

\

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- 0.2 0 i i 77 i i i ? --T - . , , i ( , , -a f?i , , i-2 2 -Jun-9 0 Y i- i ie i i 29 -Jun-9 0 0

01 -Jun-9 0 0 8-Jun-9 0 15-Jun-9 3 Date Erdmann Brook Franks / Quarry Conf.

- Precipitation

li \1ji\

C 9. ~ 0 = &. sk eo DC-8 6 4 2, 8 6 4, 2 1 1 1 1 1 0 0 0 0 O f

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March 1990 Average Flow Rates and Daily Preciaitation 2

45

- 1.8 40 -

- 1.6

~

35 ,

- 1. 4 C D

/

O 30 - O

O j.,  !

/

'\ - 1. 2 -

( v '\y - .9

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l C ,

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g N /- h l 73 is- -06 5r-

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$ 1 0

i i 16 -ucr -9 0 2 4 -uar -9 0 0 5 -vor -9 0 Date M Erdmann Brook Franks / Quarry Conf.

x Daily Precipitation BN --