Text

Semiannual Radioactive Effluent Release Rept,Jul-Dec 1989
	ML20033E231
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/31/1989
From:	Crawford A; PUBLIC SERVICE CO. OF COLORADO
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	P-90060, NUDOCS 9003090433
	Download: ML20033E231 (63)
	v • d • e

{{#Wiki_filter:-, 1 [r Public Service' ON P.0, Box B40 Denver CO 80201 0840 i A. Clegg Crawford Vice Presiderst Nuclear Operations February 22, 1990 Fort St. Vrain Unit No. 1 P-90060 iL U.S. Nuclear Regulatory Commission i Attn: Document Control Desk F Washington, DC 20555 Docket No. 50-267

SUBJECT:

Semi-Annual Radioactive Effluent Release Report f Gentlemen: Attached please find the Semi-Annual Radioactive Effluent Release Report for the Fort St. Vrain Nuclear Generating Station, i This report covers the period July 1,1989 through December 31, 1989, r and is. submitted pursuant to Section 7.5.1.e of the Fort St. Vrain Technical Specifications. Please contact Mr. M. H. Holmes at (303)480-6960 if you have any j questions regarding this report. Sincerely, h,Do i A. C. Crawford, l Vice President, Nuclear Operations ACC:SC/bw Attachments l: l ? k l "?f30sE0?Rg;p%a, R'V[ \\\\ [

c l - i b Public Service- % ' %.e l P.O. Rom 840 Denver CO 80201 0840 l 1 A. Clegg Crawford l Vice President i . Nucleer Operations 2 February 22, 1990 Fort St. Vrain Unit No. 1 P-90060 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Docket No. 50-267

SUBJECT:

Semi-Annual Radioactive Effluent Release Report Gentlemen: Attached please find the Semi-Annual Radioactive Effluent Release Reoort for the Fort St. Vrain Nuclear Generating Station. This report covers the period July 1,1989 through December 31, 1989, and is submitted pursuant to Section 7.5.1.e of the Fort St. Vrain Technical Specifications. Please contact Mr; M. H. Holmes at (303)480-6960 if you have any questions regarding this rsiort. Sincerely, gZ" & A. C. Crawford, Vice President, Nuclear s Operations Liconc!nc R:C"! Oy:.._.W @1 & ACC:SC/bw c Attachments . p 7o V m.

7-c i P-90060 February 22, 1990 cc w/ attachments: Regional Administrator, Region IV U. S. Nuclear Regulatory Commission i Attn: Mr. T. F. Westerman, Chief Projects Section B 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Mr. Robert Farrell Senior Resident Inspector NRC Office Fort St. Vrain Dottie Sherman, ANI Library American Nuclear Insurers The Exchan0e, Suite 245 270 Farmington Avenue Farmington, CT Mr. Robert Quillen Director Radiation Control Division Colorado Department of Health 4210 E. lith Ave. [ Denver, CO 80220 Mr. Milt Lammering, Chief Regional Representative, Radiation Program U. S. Environmental Protection Agency Region VIII 999 18th Street, Suite 500 Denver, CO 80202-2405 Mr. Farrell D. Hobbs Environmental Management Rockwell International Building 250, P.O. Box 464-Golden, CO 80402 Dr. James E. Johnson Radiology & Radiation Biology Dept. 135 BRB Colorado State University Fort Collins, CO 80523 4 E a

7 e 'y ,_4' pn ysp;,, g g f .1. 1 i 1 4 P ( r D, 4 u Y 1 L i t n, I a F SEMI-ANNUAL RADI0 ACTIVE EFFLVENT-RELEASE REPORT b~.- July - December i i s. ['<t 1989 -i b lin; l Y, l g ' (-?.. pf - .f <,3 ? ='9 3 y; +- s Public Service Company of Colorado = Il 't t 6,' . Fort St.>Vrain. g Nuclear Generating Station j r; "m February, 1990 x I h (- i si .f i i e c. ,.N U! l 0 5 'l, j q t i. s i)# j* i d,', t q

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}V' Attachments .i Reviewed By: Dabet; d M 3/W' I j A s. s - -1 ; -?- 4 k L il. j

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i SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT

' July - December i 1989 + r Public Service Compar.y of Colorado Fort St. Vrain Nuclear Generating Station February, 1990 i i ? ? ..+-

P-90060 February 22, 1990 1.0

SUMMARY

This report summarizes radiological effluent released from the Fort St. Vrain Nuclear Generating Station for the period of July through

December, 1989.

This information is provided pursuant to the requirements of Sections 7.5.1.e. 8.1.1.g 8, 8.1.2.d, e and j, 8.1.3.e and f, and 8.2.1.h.1 of the Fort St. Vrain Technical Specifications. This report uses the reporting format recommended by Regulatory Guide 1.21 as well as the requirements of the aforementioned sections of our Technical Specifications. The following t6 oles with a supplemental information section are included with this report: Table Description 1A Gaseous Effluents - Summation of All Releases IC Gaseous Effluents - Ground-Level Releases 2A Liquid Effluents - Summation of All Releases 28 Liquid Effluents 3. Solid Waste and Irradiated Fuel Shipments 4A Hourly Meteorological Data Please note that Table IB (of Regulatory Guide 1.21) has been omitted from this report because all of our gaseous effluents are assumed to be ground-level releases as opposed to being elevated releases. Fort-St. Vrain Technical Specifications apply exc'usively to the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-S8, Co-60. Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emnasions. This list does not mean that only these nuclices are considered. Other gamma emitting nuclides that are identifiable, together with those of the above nuclides, are-analyzed and included in this report. Sample activities that are less than the detection capabilities of our equipment are entered in this report using the value resulting from the calculation of the lower limit of detection (LLD) or minimum detectable activity (MDA). This results in reporting upper limit values that are in excess of true activities.

p e o I i i i P-90060 February 22, 1990 The lower limit of detection (LLD), for the purpose of this report, is defined as the smallest concentration of radioactive material in a f; sample that will yield a net count, above the system background, that 2 .will be detected with a 95% probability of being correct and only a 5% probability of falsely concluding that a blank observation represents a real signal. The LLD values specified in our Technical Specifications are as follows: l' Liquid L Principle Gamma Emitters 5.00E-07 uti/ml Dissolved Noble Gases 1.00E-05 pC1/ml Tritium 1.00E-05 pCi/mi Iodine-131 1.00E-06 pC1/ml i Gross Alpha 1.00E-07 vC1/ml Strontium-89, 90 (Composite) 5.00E-08 pCi/ml Gaseous Principle Gamma Emitters 1.00E-04.vCi/cc (Gas) Principle Gamma Emitters 1.00E-11 vC1/cc (Particulate) Tritium (Gas) 1.00E-06 uti/cc Iodine-131 -(Ch& rcesi) 1.00E-12 vC1/cc Gross Alpha (Particulate) 1.00E-11 pCi/cc Strontium-80, 90(Particu1&te) 1.00E-11 pCi/cc Gross-Deta (Particulate) 1.00E-11 vCi/cc Where applicable, we have listed "less-than" values for those nuclides listed specifically in eur Technical Specifications. These "less-than" values were celeviated using the observed LLD values and the total volume of the redia. The "less-than" values were not included in the total yelves for the pathway.

V i T I P-90060 February 22, 1990 The percent of Technical Specification limit on Table 1A is blank in j some cases because this value could not be calculated from data which were at or below the minimum detectable activity. On Table IC, the . continuous release mode values are not reported because this release pathway is the same as the batch mode. All other blanks on Tables 10 and 22 accur because no LLD values for these nuclides are required to be calcolated per Technical $pecifications. There has been some confusion in the past as to the total volume of water used for dilution of radioactive liquid effluent. All average diluted concentrations are based on the activity at the unrestricted

area, Although this effluent could eventually reach one of two rivers (St. Vrain-Creek and South Platte River) which converge approximately one and one half miles downstreaai of the plant, no further dilutions were assumed. Additional discussion on river flow is contained in section 4a of the Supplemental Information Section.

i On August 26, 1969, the meteorological tower was struck by lightening and consequently was inoperable until October 20, 1989. An r additional summary of data for that period has been added to this report. The data is provided by NOAA. Stability Category G (worst case) was assumed. During this reporting period, five abncrmal radioactive gaseous waste I releases were made. Four of these releases were controlled releases made from the Primary Coolart System to the Reactor Building Exhaust Ventilation System utilizing i ne Helium Purification System to purify the gas to the maximum extent possible prior to release. One release was made by venting the contents of the low pressure bottles of the i Helium Storage System as a controlled release. A total of 3.27 E+0 Ci was released during these releases. During this reporting period, trace amounts of I-131 were detected in the Reactor Building Exhaust Ventilation System silver zeolite iodine cartridges for the period commencing August 22, 1989. Iodine has not been detected again commencing September 26, 1989. These cartridges are analyzed weekly for iodine and the weekly analysis results were used to determine the amount of radioactivity released. This release path is considered to be a continuous, normal release during the affected period. A total of 1,72 E-5 Curies of I-131 was released during this period. l-There were no unplanned radioactive liquid waste releases made from the Liquid Waste System 62 or the Reactor Building Sump during this reporting period.

f i h j P-90060 February 22, 1990 There were no' shipments of radioactive material for disposal this reporting period. A change notice (CN-2939) was initiated the last half ~of 1989 affecting the radioactive gas waste system. A list of the changes per Technical Specification requirements are listed below. 2. A summary of the evaluation that led to the determination that the change could be made in accordance with 30CFR, Part 50.59. 6 'This modification was considered safety related, was not safety significant, did not require a change to any Technical Specification, and did not involve or create an unreviewed safety question. 2. Sufficient detailed information to totally support the reason for the change' without benefit of additional or supplemental information; See Attached Documentation 1. Safety Evaluation, CN-2939, pgs. 2.0-2.6, 3.0. 2. Design Input Requirements, CN-2939, B1-Bl.2, B2.1, B2.2. 3. DCAR 1370 pgs. B4-B4.2 4. Safety Related Design Analysis, CN-2939, pgs 85-B5.2. 5. Radiological /ALARA Analysis, CN-2939, pgs. B7-B7.2. 6. PPC-89-2519, August 3, 1989, Primary Coolant Activity After Final Reactor Shutdown For Defueling, pgs. B8-B8.1. In addition, the normal discharge of the Purge Vacuum Pump is to the low activity inlet header of the Gas Waste System (System 63). This . discharge is usually of sufficiently low activity levels and flow rates that they can be released directly to the Reactor Plant Exhaust System for disposal. (FSAR, Sec. 11.1.2.3)

P-90060 February 22, 1990 3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems. A modification was made that will provide a flow path [ for the Purge Vacuum Pumps (C-1301 and C-1302) discharge directly to the Reactor Plant Exhaust System (System 73). The normal discharge path of the Purge Vacuum Pumps is to the low activity header of the Gas Waste System (System 63). This change would decrease the time required to pump down the Fuel Handling Machine (FHM) or Auxiliary Transfer Cask (ATC) and expedite the defueling process. The discharge of the PVP's would be directed to the Reactor Plant Exhaust System (System 73) by closing V-1311 and V-1348 to the Gas Waste System and opening either V-13155 or V-13156, manually operated ball t valves added by the modification. The line then taps into a System 73 duct and directed through C-7309, Fuel Storage Well Emergency Booster Fan, to either F-7301 or i F-7302, Reactor Plant Exhaust Filters IA or IB, through C-7301 or C-7302, Reactor Plant Exhaust Fans 1A or IB,- pass the Reactor Plant Exhaust radiation monitors, to i the Reactor Plant i.xhaust Stack. The capability for returning the PVP discharge to 6 System 63 still remains by opening the two existing valves to System 63 and closing the two new ball valves to System 73. This would have to occur upon receiving a Reactor Plant Exhaust Stack monitnr alarm. The modification bypasses.the gas waste filters and the automatic-diversion of gas to the gas waste vacuum

tank, should activity exceed preset

limits, and discharges to the Reactor Plant Exhaust HEPA filters.

Therefore, manual action is required to direct flow back to System 63 should a stack monitor alarm occur. This change only affects the point of interface between the pucge vacuum pumps and the Reactor Building Exhaust Ventilation System, there is no change in the predicted or expected maximum exposures to individuals in the unrestricted area and general population that i differs from those previously estimated in the license application and amendments thereto. Additionally, thin change did not affect the predicated or actual releases of radioactive materials either prior to or during the change, nor does this change affect the exposure to plant operating personnel. 1

L. P-90060 February 22, 1990 There were no changes to the Process Control Program (SUSMAP-3), Issue 2, effective date November 13, 1984, during thi s reporting period. Changes to the Offsite Dose Calculation Manual (ODCM), SUSMAP-2 which were described in the previous semi-annual report were incorporated t into Issue 17 on October 11, 1989. During this reporting period, a note was ad#d in step 4.2.1 a) prior to the calculations stating, l "The allowable batch release rate 'r' is calculated using the reletse rate formula as determined in the Health Physics procedure". This was necessary to correct the batch release formula to agree with ELCO 4.8.1.la. Gaseous effluent activity monitors, or their associated recorders, inoperable for more than thirty days ending during the period. July 1, 1989 to August 14, 1989, were required to be reported per Environmental Limiting Conditions for Operations (ELCO) 8.1.1.g. None of the gaseous effluent activity monitors, or their recorders, were inoperable for more than thirty days. Gaseous effluent flow measurement instruments were also required to be reported if inoperable over thirty days ending during the period July 1, 1989 to August 14, 1989 per ELCO 8.1.1,g. None were inoperable over thirty days. On August 14, 1989, Amendment No. 71 to the FSV Technical Specifications became effective which revised.ELCO 8.1.1.g, deleting the requirement to report gaseous effluent activity monitors or flow instruments inoperable over thirty days in this Semi-Annual Effluent Report. The reporting requirements were changed to require a Special Report if less than the minimum required gaseous effluent activity monitors are operable for over seven days. It has not been necessary to issue any such Special Reports since the requirement became effective. Liquid effluent activity monitors inoperable for more than thirty days are required to be reported per ELC0 8.1.2.d and ELCO 8.1.3.e. 1 i l e

p i L i i P-90060 February 22, 1990 L One of the two radioactive liquid waste effluent monitors, RT/RIS-6212, was discovered to be inoperable on August 10, 1989 because it did not respond to activity known to be present in release number-1285. The redundant liquid waste monitor did respond as expected to the activity. RT/RIS-6212 was not returned to service until September 25, 1989. The monitor responded correctly to its check source both before and after release 1285, and it passed the 1 test procedures in effect at the time after the release. It was determined that the calibration procedure in use then did not insure that gamma photons with energy as low as 81 kev would be seen by the monitor. The activity in release 1285 consisted of Xe-133, which emits gamma photons at 81 kev. Improved calibration procedures were developed using RT/RIS-6212 which is the reason for the delay in returning it to service.- + The recorder associated with the liquid effluent activity monitors, i reportable under ELCO 8.1.2e and ELCO 8.1.3f, was not out of service over thirty days. Radiation doses resulting from the release of radioactive liquid and gas effluents from Fort St. Vrain during 1989 are reported below. Radiation doses are calculated in accordance with the Fort St. Vrain Offsite Dose Calculation Manual (SUSMAP-2), which is based on NUREG-0472, " Radiological Effluent Technical Specifications for PWR's", NUREG-0133 " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", and other inputs from the Nuclear Regulatory Commission. ~ Doses are calculated for a_ hypothetical " maximum" individual present at-all times of the year at the Exclusion Area Boundary (EAB) in the sector in which all of the activity is calculated to have been released. The following exposure rutes are for the 1989 calendar year. Liquid - 10CFR50 Whole Body 1.29E-01 mrem Bone 8.51E-03 mrem Liver 1.33E-01 mrem Thyroid 1.21E-01 mrem Kidney 1.25E-01 mrem Lung 1.22E-01 mrem GI 1.21E-01 mrem

g., I r I- ?- I' P-90060 February 22, 1990 b f. Gaseous - 10CFR50 L [ Noble Gas Gamma 6.31E-01 mrad wF' Beta 1.46E+00 mrad 1 [ Iodine, Particulates. Tritium L k ' Adult Whole Body 1.71E-01 mrem L Organ (maximum) 1.79E-01 mrem Bone 1.57E-05 mrem Teen [ Whole Bod / 1.97E-01 mrem OrDan (maximum) 2.07E-01 mrem . Bone 2.58E-05 mrem i Child Whole Body 2.82E-01 mrem Organ (maximum) 3.02E-01 mrem Bone 6.02E-05 mrem Infant Whole Body. 2.20E-01 mrem Organ (maximum) 2.64E 01 mrem Bone 1.14E-04 mrem Gaseous - 10CFR20 Iodine,- Particulates, Tritium 2.82E+02 mrem All doses are within acceptable limits in accordance with 10CFR20 and

10CFR50, It is felt that use of actual dilution factors is more accurate than i

the. annual average X/Q of 1.37E-06 s/m3 as previously reported in the Final Safety Analysis Report. Beginning in 1987, dilution factors for actual periods of release were used to calculate doses from gaseous effluent releases, , As mentioned earlier, the doses reported here are to the hypothetical most exposed. member of the public. In order to assess the actual dose to the likely most exposed member of the public due to their activities inside the site boundary, the following assumptions are made: 1) Residents living within the site boundary are at home during .50% of gaseous effluent releases.

7 I P-90060 February 22, 1990-I 2) As game fishing is not prevalent within the site boundary, fish consumption is 25% of the adult fish consumption of 21 kg/yr as listed in SUSMAP-2. [ 3) All other assumptions of SUSMAP-2 remain valid. The doses demonstrate conformance with the exposure limit in 40CFR, Part 190 of 25 mrem to the total body, 75 mrem to the thyroid, and 25 mrem to any organ, l. To show conformance with 40CFR190 subpart B, the total curies of Krypton-85 released from Fort St. Vrain is less than 9.75E-01 C1. The 29.78 key iodine-129 peak is below the minimum detectable energy of our detectors. Although small amounts of iodine-131 were released during the reporting period, the total curies were less than 40CFR190 subpart B limits. It is assumed that any iodine-129 (fission yield t of 0.574%) was well below 40CFR190 subpart B limits. The total release value of gross alpha is listed in Table 2A Section D. An annual land use census is performed.as part of the Fort St. Vrain Radiological Environmental Monitoring Program. Changes made to environmental sampling locations as a result of the annual land use census are reported in the annual Radiological Environmental Monitoring Program report. a 4

I Effluent and' Waste Disposal Semiannual Report Supplemental Information i. l Facility Fort St. Vrain Nuclear Generating Station Licensee Public Service Company of Colorado I rL 1. Regulatory Limits i L All results of radioactivity analyses of gaseous and liquid effluent -are used in accordance with the methodology and parameters listed in the Offsite Dose Calculation Manual (SUSMAP-2) to assure that the concentrations at _the point of release are maintained within the limits set forth in the Technical Specifications. These limits will ensure the quantity of radioactive effluent released from the plant is maintained as low as reasonably achievable and in any event within the limits of 10CFR20 and in accordance with 10CFR50. The air dose due to noble gases released in gaseous effluent at the unrestricted area is limited to: a) 5 millirads gamma and 10 millirads beta during any i calendar quarter, and, b) 10 millirads gamma and 20 millirads beta during any y calendar year, The dose to a - member of the public due to I-131, tritium, and radioactive particulates with half-lives longer than eight days in gaseous effluents will be limited to: a) 7,5 millirems to any organ during any calendar quarter, i

and, b) 15 millirems to any organ during any calendar year.

The dose rate due to radioactive gaseous effluent is limited to the following: a) For noble gases, less than or equal to 500 m1111 rems L per year to the total body and less than or equal to 3000 millirems per year to the skin, and, b) For I-131, tritium, and radioactive particulates with half-lives greater than eight days, less than or equal to 1500 millirems per year to any organ. l} t

.g. r The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released to unrestricted areas are limited as follows: a) During any calendar quarter to less than or equal to g' 1.5 millirem to the total body and to less than or equal to 5 millirems to any organ, and, b) During any calendar year to less than or equal to 3 millirems to the total body and to less than or equal to 10 millirems to any organ. 2. Maximum Permissible Concentrations Aii MPCs used in determining allowable release rates from the gas waste holdup system and the liquid waste system are those listed in Table II, Columns 1 and 2, respectively, of Appendix B to 10CFR20.- In addition, for the MPC of dissolved noble gases in liquid effluent the value of 2.00E-04 microcuries per milliliter was used. For the purposes of calculating allowable release rates, the MPC for halogens and particulates with half-lives longer than eight. days will be reduced by a factor of 700 from their listed value in Table II, Column 1, of Appendix B to 10CFR20. 3. Average Energy The average energy (E-Bar) of the radionuclide mixture in releases of fission and activation gases is.not calculated or ^ used 'at this facility.

4. -Measurements and approximations of Total Radioactivity a)

Fission and Activation Gases Batch releases from the gas waste holdup system are performed after sampling and analyses for noble gases and tritium. These analytical results are used along with atmospheric dilution factors to determine the allowable release rate. Gas is released on a continuous basis through a gas waste header which is monitored by a noble gas monitor and an iodine monitor. In the event of high activity in the continuous release, header control functions are initiated which divert the gas to the gas waste holdup system. j I i l l I

K i 3 i i g i All radioactive gases are released to the Reactor Building exhaust ventilation system which has a flow rate of i-approximately 30,000 cubic feet per minute. The full-flow of this exhaust is directed through high efficiency particulate filters (HEPA) and activated charcoal beds prior i to the release to the environment. b Downstream of the activated charcoal beds the gas stream radioactivity is continuously monitored and recorded by noble gas monitors, particulate monitors, and icoine monittors. b) Iodines l For gaseous iodine, the Reactor Building exhaust ventilation is monitored and recorded on a continuous basis. The 2-inch iodine cartridges used in these monitors are removed from j service after one week of service and quantitatively analyzed on a gamma spectroscopy system. The quantity of radiciodine released during that period would be calculated based on the integrated flow during the collection period. c) Particulates As in the case of iodine discussed in b) above, a 2-inch particulate filter is removed and analyzed each week, t Gross-beta analysis as well as gamma spectral analyses are s performed to identify and quantify any radionuclides. The quantity of any radionuclides on this filter with half-lives greater than eight days would similarly be. correlated to total filow during the collection period. d) Liquid Effluents All liquid effluent discharged from the site reaches the unrestr,icted area at the Goosequil Ditch. From that point the effluent can be diverted to the St. Vrain Creek via the St..Vrain Slough, or, more commonly diverted to the Goosequil Pond which is approximately one mile north of the i plant site. Outfall from the Goosequil Pond reaches the South Platte-River. Both rivers converge approximately 1 1/2 miles from the plant site. The average stream flow reported t in section Sa of this supplemental report is a summation of both rivers and was received and tabulated from data provided by the Colorado Department of Natural Resources in Greeley, Colorado.

e

i Liquid effluent is released from the site using both a continuous and batch mode. The continuous mode (automatic discharge mode) is used on the Turbine Building Sump effluent where the only expected radionuclide is tritium.

This discharge path utilizes a continuous sampler and an aliquot of this composite sampler is taken three times per week and analyzed for gross-beta, gross-alpha, tritium, and gamma emitters. Total flow integrators enable us to calculate the total activity released via this pathway based on composite sample results. Discharge from the Turbine 1 Building Sump is made directly to the unrestricted area with t no dilution. The batch release mode is used on the Reactor Building Sump effluent and the liquid waste processing system. The Reactor Building Sump area can hold several hundred thousand gallons of waste water from various sources which could be contaminated. The liquid waste system consists of 2-2000 gallon receivers, 1-2000 gallon monitoring

tank, and l

associated filters and demineralizers. This system is designed to collect and process contaminated waste water resulting from reactor and laboratory operations. Prior to each liquid batch release duplicate samples are quantitatively analyzed for their radioactive constituents. These. analyses include gross-beta, gross-alpha, tritium, and gamma spectral analyses. The results of these analyses, and other analyses as dictated by the gross-beta results, are used to determine the maximum release rate from the site. The liquid effluent, normally released at or less than 60 gallons per minute, is diluted with cooling tower blowdown, which runs at or core than 1100 gallons per minute. The resulting mixture is sampled during the release period to t confirm compliance with regulatory limits. l The liquid effluent from the batch release mode is monitored continuously by redundant gamma activity monitors. All tank level indiceting devices, flow monitoring and recording devices, and radiation monitoring equipment are calibrated and maintained at scheduled intervals in accordance with established procedures.- L Composite samples from batch releases and continuous releases are analyzed monthly for Sr-89 and Sr-90. Composite samples from batch releases from the liquid waste processing system are analyzed monthly for S-35. All sample results are conservatively decay-corrected to the start of the composite period.

I 5- [ e) Overall Errors The overall error associated with determining the total activity released from the site for-both gaseous and liquid effluent is estimated to be 17.3 percent. This value is the square root of the sum of squares of counting statistics and associated calibration errors, sampling errors, and tank volume estimates, each considered to be plus or minus 10 percent. 5. Batch Releases a) Liquid l l l JNumberofBatchReleases 117 i llotal Time Period for l Batch Releases 1.04E403 HOURS iMaximum Time Period for la Batch Release l 4.65E+01 HOURS l Average Time Period for la Batch Release 8.92E+00 HOURS l Minimum Time _ Period for la Batch Release 1.5BE+00 HOURS I l Average Stream Flow During l 'l Periods of Release of l' l Effluent into a Flowing l l l Stream 1 2.50E+05 GPM* 1

Gallons Per Minute b)

Gaseous I 1 Number of Batch Releases l ,49 i Total Time Period for l JBatch Release 1 2.22E+02 HOURS l Maximum Time Period for l Batch Release i 1.62E+01 HOURS I l Average 1 4.52E+00 HOURS l l IMinimum i 1.40E+00 HOURS I i 1 I

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TABLE 1A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1989 GASEOUS EFFLUENT-SUMMATION OF ALL RELEASES l Unit l Quarter l Quarter l Est. Total l l l 3 l 4 l Error. % l i l l l l l A. Fission and activation products i L 1. Total release l Ci l 9.11E+01 5.13E-01 l 1.73E+01 l L l l __ 1.15E+01 I .6.45E-02 l l 1 E .P, Average release rate for luci/secl period l l l l l l 1._ l r P 3. Percent of technical l l 1.75E+00 l 3.63E-02 l specification limit l l l l B. Iodine 4 1.72E-05 i <3.58E-06 l 1.73E+01 l 1. Total iodine-131 l C1 l 1 l l l 2. Average release rate for luci/sec. 2.17E-06 l <4.'51E-07 l period l l l 1: l l l 3. Percent of technical l l ~ 2.41E-08 l specification limit l l l l C. Particulates 1. Particulates with half-lives l Ci l <1.41E-07 l 1.51E-07 l 1.73E+01 l > 8 days l l l l l l l l l l 2. Average release rate for luci/secl <1.77E-08 l 1.90E-08 l period l l l l 1 l 1 l 3. Percent of technical l l l 3.17E-10 l specification limit l l l l l l l l 4. Gross alpha radioactivity l Ci l <4.19E-08 l 3.55E-08 l 1 l l l D. Tritium l '. Total release l Ci l 1.86E+00 l 1.60E-01 l 1.73E+01 l l l l l l 2. Average release rate for luci/secl 2.34E-01 l 2.01E-02 l period l l l l 1 1 I I l 3. Percent of tech spec. l l 3.90E-03 l 3.35E-04 l i; limit l l l l

try p, 3, TABLE IC-W y EFFLUENT' AivD WASTE DISPOSAL SEMIANNUAL REPORT (1989) y. t i -.GASE0tlS EFFLUENTS--GROUND-LEVEL RELEASE [j. CONTINV0VS MODE. BATCH MODE r id" 1 Released

l. 0 nit j.

l l Quarter.3 l Quarter 4 l' E l' l. l l l-l l-( F_ission gases h lUypton-85 i C1 l l l 8.43E-01 l 3.24E-01 1 ~ l: l-l l l l 1 .-l krypton-85m l C1 I l l 3.51E-01-l l l 1 I I I I I l .. :.'l krypton-87 l C1 l l l 6.44E-02 l <6.10E-03 1 l i I I I I I ..j krypten-88 l Ci l l l -3.29E-01 l <8.81E-03ll

li l-1 I

l l 1 ,l xenon-133 i Ci l l l 8.62E+01 l 8.10E-02 l l~ l l I I l l' -l xenon-135 i C1 l l I 2.35E+00 l <2.49E-03 l 3 I I I I I I I ~ lxenon-135m l Ci l l l 5.52E-03 l l -1 l l l 1 l l-1 lxehon-138 l Ci l l l <9.54E-02'l.<2.92E-02 l l-l __.C i I i i I l l l l l 7.41E-01 l <2.09E-02 l -l xenon-133m l' I I I I I u l xenon-131m l Ci l l-l 2.06E-01 l -1.08E-01 L l l l_ L_ L I-I-~ Ci l l l 9.11E+01 l 5.13E-01-l l Total for period

l

.l-l l l l l .__ l -Iodines liodine-131

l. C1 l

l l <4.20E-10 l <1.28E-10-l l 1 I I I l -l c _l1odine-133 I Ci l l l l l-l- l l l l l l -liodine-135 l - C1 l l l-l l L1-l l l l l l Total for_ perica

l-Ci l

l l l -l- ~ l l I I l l l NTotal' values do not include "<" data (

yygm s; o. '~ .m;;

n e

1 ,ym

1

_a rt; J.u;c'p - *. f u. s* ~ me i y 3 4., e e v,s ,L ~y> 1 s g s [ A S 4 5 a.. _ pyw TABLE 3C'(Continued) . Page_2- [e3

'(

a& : t

0'f L t

(; f ' ' (' f D'? .' t ? ? w +

g; ParticulatesJ zi

( O strontium-89;

1 Ci l

I >l -l 1 x. i -l-l- l.. ? ,a- ', V: I stronti um-90,- - l x-c1 l-I l.. 1 a 4 F: l l. l-l l. l' ' l- 'A l Ci l' l~ l .l .l; cesium-134 ~. 1 N c l-l l l ~l-l.; 4 l cesium-137: I ci-l-l .l. 1 4

l I:

<l I I n barium-lanthanum-140 l. Ci-l: l-l l. ~ 'l .l' l l' - l -- i n 8 s a u .p, y a c p If s 5 5. 1 A-3 A 7 '4 ., { 4- [ . v_ .t %;y s s t . x 'I r ,Q: ;,.,' ,Q e d

o.

,[ , i'. e

Sh Q ', p, _

sc ;a, j_ ,f. w - + t it f ti 1 w.... 1 i,' VSUTota,l,"valu'es do not include '<" data-i]//? t

r 4

i a . ; i s.'._ M g %.d [.fj I e l.t. s r se,. e r ~O;' h,. _k $$ a

g-TABLE 2A p 1 [ EFFLUENT AND WASTE DISPOSALLSEMIANNUAL REPORT 1989 ~ ~ LIQUID EFFLUENT-SUMMATION OF'ALL RELEASES l Nuclide l Units ! Quarter j. Quarter l Est. Total-l l l l 3 l 4 l Error. % l 1 I I l l l A; F1ssion-and activation products. l = 1. Total release lCs-137 l-Ci l 2.13E-06 l 1.45E-06 l 1.73E+01 l l l l l l l t

2..

Average diluted luc 1/ml l 3.86E-12 l 2.63E-12 l concentration l. l { l I I I l 3. Percent of-l l 1.93E-05 l 1.32E-05 l -applicable limit l l-l l 1. Total. release 10o-60 l Ci l 0.00E+00 l 6.49E-07 l 1.73E+01 -l l l 1 l 1 -l 2. Average diluted luci/ml l 0.00E+00 l 1.18E-12 l concentration l l l l I l l 1 3. Percent of l l l 3.93E-06 l applicable limit l l l l B. Tritium 1.~ Total release lH-3 l C1 l 3.62E+01 1 3.64E+01 l 1.73E+01 l. l-l l l l l 2 ;-- Average diluted luc 1/mi l 6.56E-05 l 6.60E-05 l . concentration l l l l 3; Percent of. 2.19E+00 l 2.20E+00 -applicable limit l l l l 'C. Dissolved and entrained gases 1. Total release lXe-133 l Ci l 5.41E-02 l 5.87E-06 l 1.73E+01 l. I l i l l l 2. Average diluted luCi/mi l 9.81E-08 l-1.06E-11 l concentration l l l l I I I I 3. . Percent of l l 4.91E-02 l 5.30E-06 l

c applicable limit l

l_ l l L g;f A

?N,1 y Page 2-TABLE 2A(Continued) 1.- Total' release ~ lXe-133M I Ci l 2.72E-04 l 0.00E+00 l 1.73E+01 1 F l l l l l-l- D 2. Average diluted lMC1/mi-l 4.93E-10-l 0.00E+00 l concentration-l l l l' I l-l l 3. Percent'of I l 2.47E-04 l l applicable limit l l 1 l n. 1. -Total release lXe-131M l Ci l 8.12E-04 l 0.00E+00- l 1.73E+01 l 2. Average diluted l l 1.47E-09 l 0.00E+00 l ~ ~ l 1 l l l l concentration-I l-l l l l l 1 3. Percent of. l l 7.35E-04 l l

applicable limit l

l l l .D._ = Gross alpha radioactivity 1. Total release-l Ci !<1.76E-05 1 <1.95E-05 l 1,73E+01 l l l l l l .'E Volume of.' waste released -(prior to dilution) -l Liters i 1.58E+07 l 1.27E+07 l-1.00E+01 l I I I l-l ' F.- Volume of' dilution water used during release-l Liters l-5.52E+08 l -5.52E+08 l 1.00E+01 1 7 I l l l I w ,V i ..) f a iN.

L n; TABLE 2B-1 o l EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1989) LIQUID EFFLUENTS [ CONTINU0US MODE BATCH MODE P INuclides Released i Unit I Quarter 3 l Quarter 4 I Quarter 3 l Quarter'4 I _ l-l l l-l l l- - I stronti.um-89 i C1 l <8.27E-05 l <2.05E-05 I <1.40E-05 l <7.54E-06 l. l' l 1: l l l l I strontium-90 l C1 l <4.70E-05 l <2.11E-05 l <1.4'E-05 1 <7.78E-06 l l' i l l l l l 4 1 l cesium-134 i Ci l <7.41E-04 l <6.52E-04 I <3.64E-04 l.<2.37E-04 I l I I I I I l-cesium-137 l C1 l <7.07E-04 I <6.22E-04 l <3.24E-04 I <2.08E-04 l l l l l 1 I I iodine-131 l Ci l <6.18E-04 11 <5.43E-04 l <3.03E-04 l <1.96E-04 l l-1 I I I l-l 11 cobalt-58 l Ci l <6.58E-04 l <5.80E-04 l <3.23E-04 l <2.10E-04 l - I l l I I I I cobalt-60 I C1 I <6.34E-04 l <5.58E-04 l <3.12E-04 l <1.87E-04 l l l l 1 1 I I 1ron-59 I Ci l <1.47E-03 i <1.29E-03 I <7.21E-04 I <4.68E-04-1 I I I I I I i -I zinc-65 1 Ci I <1.50E-03 l <1.32E-03 i <7.36E-04 l <4.78E-04 I I I-1 I I 'l 1 - l-manganese-54 l C1 i <7.13E-04 l <6.27E-04 l <3 50E-04-I <2.27E-04 l -I l l l I I l l-- chromium-51 l Ci 1 l l l l l l l l 1 _ l l + il zirconium-niobium-95 l-Ci l 1 l l l-l -l I I l-1 I l ~.I molybdenum-99 I C1 I <4.80E-03 I <4.22E-03 I <2.35E-03 I <1.53E-03 I d l I -l l I I I .l . technetium-99m l Ci l l l l. I l .I I I I I I I .j-barium-lanthanum-140-l Ci I l l l l l l -l l I l l I I I, cerium-141 I C1 l <9.96E-04 l <8.77E-04.l <4.89E-04 i <3.18E-04 l l. I I I I I l l-tritium l-C1 1 5.45E-01 l 1.62E-01 l 3.56E+01 1 3.62E+01 l =l-1 I I I I I I sulfur I Ci l <2.27E-03 l <1.51E-03 l <1.18E-03 I <8.58E-04 l )

l '

l-l- _l l l 1 1-Total.for= period (above)*l C1 1 5.45E-01 1 1.62E-01 1 3.56E+01 l 3.62E+01 l' l' l l I l l j * Total values do not include "<" data i Ii J

{+Sf.'.~,9p ;YU h-6 __, ..s, l <m m 4 n/'- myg,.. m-w yg g vi - ae 4 's nM-,,' t "L i

b

- 1 y);4 ?., _' h't { 3 b v r r s e,,,.t -.h ;. -- -- .g. e ch e '7 7;;. ..r.: ~ Page'27 T g~, f - i u'- . TABLE'28_(Continued)( 3 t [dO,,, i f.j-1 y y 3. IN h';?.U...4~y 7 s l ^.' Continuous Mode . Batch Mode" ~ f ?,

+

pf ' pp , A. y ?l1 xenon-133! l: Ci l <2.10E-03 --l ;. <1. 85E l : <5. 51E-02_ <6.23E-04 l:- ip tl l' 'l l l: 1

l, c: 1. xenon-135 C1 l <5.54E-04 l-<4.87E-04 l <2.72E-04-l

<1.76E-04 l.': a g,-- l :w l p l l l= g f' ^.'. d,-.j N ,.i c. ( s i , H'.. -: cnt,< ,>k[4 i r n n ~ Do k '.J'd.- q,. > y T + g'.? yg _. a-. p-A i k i. ((i s f ..,j k fgh J 'w, s e . 8 s v_- p ". f s Ul .'(_'. t u iW.

n 4

e l.. 4 ....?.;'.'$3' ..Q. u f i k f lew. <s i ]-fg. e 1 [isz.W tt l l k [ [ l ]-l Jh.; s s, .-./ ):n.. ,..g + + R oz 4g : I 52. g ,\\ M;mN NTota1) values do:n'ottinclude'"<" data j n x m rm ->a y 'y 7 : o. d;; M,,

ug i

'M f(!l \\1..I ,,.;.=.- ? ~s. ,-.-h)L r 4 4- ,f ' I: 3'k.' a [ 3.h. - "

ty_

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< '_ s: 3 m

4 g ,c+j t n 1 ' c^ g_fo x g;( -

- d

'-a",, .i -] gigi Ahg.,. 4 i 4

,y l TABLE 3

EFFLUENT AND WASTE DISPOSAL SEM1 ANNUAL REPORT-(1989)-

S0' LID WASTE AND IRRADIATED FUEL SHIPMENTS . A.- SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel) l 1. Type of waste l Unit l 6-month l Est. Total l l5 l l Period l. Error. % l l l l l -l: l i I I l a. Process-waste, i..e., l m' l l l spent resins, filter l Ci l 0.00E+00 l 0.00E+00 l sludges, evaporator l-l l l 'l = bottoms, etc. l l l l 'l l I l 1 1 l l l l l- 'b . Dry active waste i.e., l m' l l l -l- ' dry compressible waste, l Ci l 0.00E+00 l 0.00E+00 l contaminated equip, etc. l l l l l l l l l l l I I l -- c. Irradiated components,- I m' l 0.00E+00 l 0.00E+00 l l: control rods, etc. 1 Ci l l -l L 1. l l i ,l l l 1 -1 -- l ~ _d.. 0ther (describe). l m' l 0.00E+00 l 0.00E+00 l- -l Ci l ,l l l l l-j' 2. Estimate; of major nuclide composition (by type of waste) Type of Waste l Isotope l Content l Curies-l Error % 1 _ ] l i 1 a. Pror.ess waste I t l l l I i 1 '1 l l I I b. Dry active waste 'I I l l l l-I- 1 1 I I I I I I I l c.. Irradiated Compo-l ,L l I nents I ,J. I I d. Other I i I l , J 3. Solid Waste Disposition Number of Shipments Mode of Transportation Destination 0.,_ _

B.

IRRADIATED FUEL SHIPMENTS (Disposition) Number of Shipments Moi 7 ..:.x - . ion Destinatien _0_ j >A

ggy >y, m,. - - ),y y . :j) f 4 P-90060 February 22,=1990 4 i s 'I i s = 1 Tabic 4A - i, s llourly Meterological Data -j 1 4 i .3 k .i i k

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my + - y q l 1 l i FRDSAflLITY Df.00@lT10N DCCURRENCE WITHIN STA!!LITY CLASS PERIOD 0F, REPDRit ' 89/08/26 21:501001 THRil B9/10/1912:00:00 = pultlifCLASS 6 r WIND' RECT 10N 0.75-3 4 B -13-12 19 24 )24' TOTAL LN 2.591E-02 2.306E-02 7.546E-03 1.353E-03 0.000E+00 0.000E+00 5.787E-02 n._ ' ' ME : 2.442E-02 3.452E-02 1.0!BE-02 1.424E-04 0.000E400 0.000!+00- 6.926E - - NE. 2.933E-02; 4.32BE-02 1.125E-02 1.566E-03 1.424E-04 0.000E+00 8.556E-02 .5 J ER '2.712E-02 .3.331E-02 8.542E-03 1.!!9E-03 0.000E+00 0.000E+00 7.012E41 x

I' 2.463E 02

!.025E-02 '8.542E 0.000E+00 0.000E+00 0.000E400 3.573E4/? ESE -1.90iE-02 1.061E-02 7.ilBE-04 7.!!8E-05 0.000E+00 0.000E+00 3.040E 02 3 1.716E-02 -1.794E-02 5.766E-03 2.136E-04 1.424E-04 0.000E+00 4.122E-02 h J5SE -1.765E-02 -2.136E-02 9.610E 4.983E-04 0.000E+00 0.000E+00 4.912E 02 { 'S! '2.242E-02' l.723E-02 4.627E-03 1.424E-04 0.000E+00 0.000E+00 4.442E-02

ESW 3.972E *435E-021 2.915E-03 2.647E-04 7.!!2E-05 0.000E+00 6.7'4E-02 3.428E 02- - 2.420E-02 _4.200E-03 1.780E-03 7.!!9E-05 0.000E+00 6.513E-02.

t b WSW 1.303E-02 6.264E-03 2.349E-03. 7.030E-04 2.047E-04 0.000E+00 2.271E i k' W. 6.3!5E-03 2.047E-03 -B.542E-04 -6.407E-04 2.!!6E-04 0,000E+00 ~1.089E i ~WNW/ 6.834E-03 3.915E-03 5.695E-04 -1.424E-04 0.000E+00 0.000E+00 1.146E-02 -1.025E-02l ?4.933E-03 1.566E-03 3.559E 0.000E+00 0.000E+00 1.716E-02 ' NWW 2.007E 1.772E-02 9.183E-03 2.27BE-03 7.118E-05 0.000E+00 4.933E-02 4 TOTAL -3.3BBE 2.952E 01' B.072E-02 1.139E-02 9.9e6E-04 0.000E+00 7.277E-01 l I4 PROB 421LITYOFChlMDCCURRENCE*2.723E-01 q Y i. n 5k ' l

ISSUE, STATUS . --4 4 gy L SY CN. NO, Public- " " ' S ' V " " W C" ' " "ePfw9tW44 "^j'. f,.,, g., o Service,- eunuc senvece comenny or cocooAoo ,y 4 i, U2r f_ % p .f J ' / Acct CHAP.in NOTICE ~ O W_dX/Ard _ [p.h4_.dI d' M I. MA? MC#e/ REASON FOR CHANGE. . depastserud-festate-rs'nol'sy a 7 31/esgr s/L. car 2921 la CN GENERATED IN RESPONSE TO ICOMMITMENT JOB NO. DCAR.1 Et E BULLETIN. ETC.): A__!f8 /870 $*#Md#^9 SYST. NO: l3 )) k',A 75 /l / EQUIP. NO: SPEC. NO. SAFETY EVALUATION: SEE DETAILED JAFETY EVALUATION X SAFETY RELATED ' ENHANCED OVALITY NONSAFETY RELATED g, MODIFICATION ^ DOCUMENT CHANGE ONLY f 4M. ARKS & EdN m n jt.y &'e M 62 k $*0 5 e S hRE .. J a u,,ep,6 6 A w k h s ^ ~ t. ,A s r ep [c. - w y [ erd b,bd* l,4* d, 5* A, O 'i,53~, d

4n 2.

t 3 [H t/r w t$ 'Dunw.el- {[{4l We APPROV AL S ON IO ENGINEERING ...aNa r un a, ioA r e. l PROD. . us T t Olat 11<9 .u.~.iu e iv. u,- Y $$49I6446%

b J v'C N N

$. 9... 0A c _ ~.... u PORC REVIEW. REQUIRED IF CH ANGE IS SAFETY RELATED OR HAS POTENTIAL FOR CREATIP b AN UNREVIEWED ENVIRONMENTAL QUESTION. YES NO . SAFETY SIGNIFICANT " UNREVIEWED SAFETY OUESTION UNREVIEWED ENVIRONMENTAL QUESTION s.osarwei .eT o e.o ca n. NFSC REVIEW; REQUIRED ONLY IF PORC DETERMINES CHANGE TO BE EITHER SAFETY SIGNIFICANT. AN UNREVIEWED SAFETY OUESTION OR AN UNREVIEWED ENVIRONMENTAL OVESTION. i YES NO l UNREVIEWED SAFETY OUESTION . UNREVIEWED ENVIRONMENTAL OUESTION .uro no oa rsi NRC REVIEW.

REQUIRED ONLY IF NFSC CONCURS THAT THE CHANGE IS AN UNREVIEWED SAFETY OUESTION OR ANUNREVIEWED i

i ENVIRONMENT AL QUESTION. l NRC REVIEW COMPLETs0 ,y, ' NRC REVIEW REQUIRED PRIOR TO IMPLEMENTATION CL OSE OUT (To Be Completed By NED) OOCUMENT REVISION COMPLETED geuN A 1 un t i de A T E CHANGE NOTICE CANCELLED F#miFi 344 30-3380

1 T' PubilC FORT ST. VRAIN NUCLEAR GENERATING STATION $9fVICO. Pusuc SERVICE COMPANY OF COLORADO CN TCR/SCR/PC/TR NO. 2?N ' SAFETY EVALUATION PAGE 1D ) ( Alf GOHV CN OVERALL 0 CNSUBMITTAL O SETPOINTCHANGE REPORT O TEST REQUEST LJ TEMPORARY CONFIGURATION REPORT O PROCEDURE CHANGE (FSAR> 0 OTHER CLASSIFICATION: ARE THE SYSTEM (S) EQUIPMENT OR STRUCTURES INVOLVEO, OR DOES THE ACTIVITY AFILECT: CLASSI $ YES O NO ENGINEERED SAFEGUARD D YES E NO SAFE SHUTDOWN NYES O NO PLANT PROTECTIVE SYSTEM O YES E NO SAFETY RELATED 3 YES C NO SECURITY SYSTEM O YES S NO REMARKS I V At U A llON Use Addition.el Sheets if Heguared L DOES THIS ACTIVITY AFFECT STRUCTURES, SYSTEMS, COMPONENTS. EQUIPMENT TESTS, EXPERIMENTS OR PROCEDURES ' DESCRIBED IN THE FSAP.OR TECH SPECS? %YES O NO LtST THE APPLICABLE' SECTIONS REVIEWED: NM bS'swf ' /e @ 8. E / E 2 _I I'f" 1 k l.2 d, E 1 2.7 9, T~ f d', 9 Lf'4L, f.$ 'f,2,f f 'f,,2,2,f/f (L22 4,2.1 ? l.$, 7. /> 2. le, 7,$,$,1, 9,l. T. f,2. ? AL. i+ 2,IJ. Isj.1. 'A /7,4.' S, a f t+,2'-r. r1.L ii + -1. if I. II.2.,4 !L)u 2 CL i o ~ Tend $b c1 ! Al.O hof. 9. 4 Y2. V,D /, t,+. 7. 7 '4. .5. // S. I, $E b 1,P.$ 7 2 'h *), [f.t" TWAf.dfl.h FPPP 2 R,. 3.,I,.tJ6 s'.' Fpd-h 'Drs 9./.2'GrR %.'l.1 2. DOES THE ACTIVITY REQUIRE THAT CHANGE (S) BE MADE TO THE FSAR OR TECH SPEC? % YES C NO LIST SECTIONS TO BliCHANGED AND THE CHANGES TO BE MADE: NIMM bb% f 9. /, k, Mo is 7 iI/ I M a.c es c roh d Sen (we r

9. I-F 9. 4 -i, 9. 7-1 a d ll f l -l' mL renjre.

rl, w a%>a Jn>Y M -He ha e'F sd L ens fi-l> Dia se r 16 w % L ce w, W, % f *Di4u'J I,w n " C A J. FPPP.frrhww 7 t a'd 9. I n,- e un w edbL' _r 4s Len.L S're. unos//d:- or. I ~ d.'L%.-l J.o YO tl. he.l' e Wdu H _t th-e.- V'd.run'r e d. 3. DETERMINE WHETHER OR NOT THE ACMdlTY INVOLVED IS AN UNREVIEWED SAFETY QUESTION UTluZlNG THE FOLLOWING (A) HAS THE PROBABILITY OF OCCURRENCE OR THE CONSEQUENCES OF AN A*CIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE FSAR BEEN INCREASED? O YES S NO STATE BASIS: ee et a f-e l (B) HAS THE POSSIBluTY OF AN ACCIDENT OR MALFUNCTION OF A DIFFERENT TYPE THAN A EVA ATED PREVIOUSLY IN THE FSAR BEEN CREATED? O YES S NO STATE BASIS: be d / l (C) HAS THE MARGIN OF SAFETY, AS DEFINEDI HE BASIS F R ANY ECH iCAL SPECIFICATION OR IN THE FSAR BEEN REDUCED? O YES % NO STATE BASIS-

1. a aC #

'DOES ACTIVITY APPEAR TO: INVOLVE AN UNREVIEWED SAFETY QUESTION O YES @ NO d BE SAFETY SIGNIFICANT O YES R NO El. /O~10 U APPROVED d 0-BY l'- 65fGasAfumo g iDA ffi (SiGesAIVRip i M" I (DKTO r DAP 10.27-94 C sorml01344433E2

n t Safety Evaluation for CN-2939 Page 2.1 R p BACKGROUND: This CN provides the design input requirements for five changes and the permanentization of TCR 88-07-02, i

1.. The first change is the addition of a new controller, PDC-1156, on Main: Control Room Panel, I-9303.

CN-2983 will install other equipment in the plant for use in controlling PCPV pressure during defueling. 2. The second change adds equipment / piping necessary to allow draining of the System 23 Front End Cooler (s) (FEC) at PCRV pressures of atmospheric or below. The-FEC will remain in service to remove moisture from the primary coolant when/if moisture content is in excess of approximately 45 F dewpoint. Major components will be a-vacuum pump, a receiver (tank)and associated. piping and valves. The pump and receiver will-be ' located on Level 110.in the Reactor Building. The pump will be used to pump down the receiver:to facilitate FEC draining. Water drained from the FEC will-be directed to.the System 61 Decontamination Tank, T-6101.- The liquid effluent will-be handled as liquid waste and released in accordance with normal release procedures (i.e., ELCO 8.1.2). _3. - The third change will provide a flow path for Purge Vacuum Pump (PVP) discharge directly'to the Reactor Plant Exhaust System (System 73). This change. would decrease the time required to pumpdown the Fuel .i Handling Machine (FHM) or Auxiliary Transfer Cask (ATC) and expedite the defueling process. The normal discharge path to the Gas Waste System (63) will remain as an option. + 4. The fourth. change will provide chilled ~ water cooling for the FHM Purge Vacuum Pumps. This modification will improve vacuum pump reliability by reducing or eliminating high temperature trips. L The chilled water will be provided by the existing chillers / pumps located on the Fuel Deck, Level 11. 5. Some electrical cables 'previously-used in conjunction with the Nitrogen Pressurization System (NPS) will be used t.s described on -CN page B12. NPS component removal was discussed in DCAR 1370 and determined to be acceptable since the equipment / instruments are no longer used or needed. 6. TCR 88-07-02 was originally installed to jumper time delay relays (TOR) on the Nitrogen Recondensor Chiller unit, S-4602, which allowed for continuous operation during low heat load situations. Also, one electrical lead has been lifted to disable one stage of the compressor which lowers the electrical load. l-

'y f-. Page 2.2 S-4602 and associated Glycol / Water Pumps, P-4605 and P-46055, will provide the cooling water to the Purge Vacuum Pumps as discussed in Change-#4 above. These modifications will be considered Safety Related, Class 1, and Safe Shutdown due to addition of instruments on 1-03 in the Control Room, cable routing, and addition of pipes / pipe supports to structural steel. The Reactor Plant Exhaust Filters, each of-which includes a moisture separator, a high ef ficiency particulate air filter (HEPA) and a 1 charcoal adsorber, will filter gas exhausted from the'PVP prior to being monitored and released to.the atmosphere. The exhaust system is designed to process radioactive gaseous ' effluents such as may be associated with MCA or DBA 2' (Maximum; Credible Accident; Rapid Depressurization/ Blowdown, respectively). The gaseous effluent activity expected from the defueling process 1 (refer to.CN -page B8) is much less than either of these accidents described in FSAR 14.8 and 14.11 The reactor plant exhaust filters are engineered safeguards' equipment. They receive and filter reactor building exhaust to remove airborne i contaminants, etc, prior to that exhaust being released.to atmosphere. H FSAR section 6.2.3.2.3 discusses the reactor plant exhaust filter design in detail. The filters are designed to handle 19000 CFM each f at up to 450 degrees F..The FEC effluent dewpoint wi11 be maintained below 50 degrees F. hich.i s below exhaust filter limits. As-w described in the CN, maintaining the PCRV' pressure. at: conditions, suitable for defueling will result in conditions within the design of the. filters. The.' Design Analysis section of the CN evaluates and demonstrates the-adequacy of the reactur building exhaust. filters' capacity to -accommodate gaseous effluent during the defueling process. Additionally, FSAR section-6.2 describes the reactor building ventilation

system, the exhaust fans and. filters (engineered safeguards), and the design bases of the sy stem.-

Gaseous. releases .through the reactor building ventilation system will be planned releases, fully controlled and monitored, which will be maintained x within the limits identified in 10CFR20 and in accordance with ELCO 8.1'.1 ~. Also, existing Health Physics (HP) practices and procedures will be used to ensure compliance with 10CFR20. In conclusion, HP is involved with the defueling process and knowledgeable of the potential radiological concerns involved with i '4 - o gaseous / liquid effluents (refer to CN-2939 pages B7 thru B7.2). The radiological /ALARA concerns are adequately addressed and found to be acceptable. i

y E I i i t Page 2.3 i Per the analysis contained-in the CN, this modification introduces no e new fire protection concerns. Reference CN-2939 pages B5.1 and B14. The vacuum pump being installed to assist draining System 23 FEC is a pump which has previously been used in the plant. Various portable vacuum ~ pumps have been used in the plant to, for instance, evacuate System 25 vacuum jacketed piping. Therefore this installation does not introduce any new fire protection concerns. SAFETY EVALVATION: 3.(A) The probability of occurrence or the consequences of-an accident or malfunction'of equipment previously evaluated in the FSAR has not been increased. There will be no change in the function or operation of the existing equipment being utilized in this modification. The' new equipment being installed will be compatible with existing -components / piping and performance requirements. PCRV pressure control will be enhanced through use of the new instruments installed in the' Control ~ Room and on 1-03. This installation is considered acceptable per CN design analyses on pages B11.3, B12, B15.1, B15.2 and B15.3. The new: equipment / piping-installed to drain the System 23 FECs will A provide a controlled, contained method to remove and discard.the potentially contaminated liquid ' removed from the reactor during PCRV pretsure control. The' plant has been shutdcsn since August-18,. 1989. Any liquid _ drained from the FEC will nave less activity than is generated during normal plant operation..Therefore, rupture of one of the' new lines would be less severe than rupture of existing FEC drain lines during normal reactor operation. Also, any major' leakage would-be. collected by the Reactor Building drain system and processed in the usual manner. Use of. flexible (rubber) hose and Chicago Couplings (metal) to supply cooling water to C-2313 is acceptable since the installation and equipment are not safety related and the hose is-designed for conditions well in excess of those expected for this system (Refer to CN page Bl.1 and B1.1.4). The new flowpath from the PVPs to-the Reactor Plant Exhaust System will^be the preferred, acceptable flowpath since the exhausted reactor helium. activity level will be known and monitored via this pathway. PVP discharge to the gas waste system remains an option, if and when

required, simply by normal valve manipulation. Accidents involving gaseous waste are discussed in FSAR 14.6.2.

Given expected primary coolant activity levels (CN page 88), rupture of a new gaseous effluent line' would be well within the limits of 10CFR20 and 10CFR100, as described in FSAR 14.6.2. .I i f g p O' - -P@ *g

s n. F4 f Page 2.4 Post. accident monitoring capabilities are not affected by this CN installatun. Samples are required to be taken from the High. x " Temperature Filter Adsorbers and at the Analytical Instrumentation-Panel. The existing compressor associated with the sampling is not as reliable as. desired,

however, as indicated on CN page B7, HP has

. initiated replacement-. of the compressor. This-will enhance-our ability to determine activity and moisture levels in gas to be released. Loss of outside electrical power (LOEP) is not a concern since the d Reactor Building Exhaust System would continue to function (connected to,Tessential buses), PVPs would cease operatinn and pump suction. valves fail closed preventing oil discharge into suction lines and' back into the PCRV. LOEP recovery is addressed in. existing plant procedures. J The provision for chilled water to the PVPs is acceptable since the ~ PVPs already use cooling water. The only difference is in the source of._ supply (existing equipment), temperature of the cooling water, and i potential-slight increase in water condensation in the PVP oil. The. lowered cooling water temperature may result in a slightly highe_r i motor starting load, but is insignificant as demonstrated -during the PCRV evacuation exercise-earlier in 1989. PVP oil condition is regularly checked during periods'of PVP operation. Neither of these effects reduce pump reliability or safety. Chilled water improves pump availability. 1 Since the Nitrogen Pressurization System'is no longer used or needed (Reference DCAR 1370 and CN page B12), various electrical cables which L run. from the helium storage building to the Control Room will be determinated. Utilization of NPS cabling will have no effect on any safety-related equipment or system, or on any equipment required for j the defueling process or continued safe plant operation. CN-2983 will. j L - address further NPS and/or PCRV pressure control equipment items. l t Permanentization of TCR 88-07-02 is in accordance with NED procedures and directives, and' enables the non-safety related chiller unit and pumps (2) to operate more reliably during periods of low heat load. i .The additional equipment and/or components added to the Reactor Building ' Fire Hazards Analyses is considered negligible per CN analysis. 3.(B) The modifications covered by this CN do not introduce the possibility. of a new type accident or malfunction. This conclusion is . based on the information provided in 3.(A) and the BACKGROUND Section, .and the CN design analyses documented in the CN. 1

c 4 x Page 2.5 Since the existing equipment, new equipment, and new piping and valves are all non-safety related, all qualified for the expected service,- -and installed in accordance with applicable codes and specifications-(refer to CN pages Bl.1'and Bl.2) no new accident modes are created. also, where appropriate.or required, new instrumentation is installed to monitor equipment and/or processes a,nd existing instrumentation-is' utilized to monitor 'potentially radioactive gaseous or liquid effluents. 'During. work in progress in the Control Room adequate precautions heve I been provided on CN page G4.7 to ensure Reactor Operator access to the Control boards, avoidance of electrical shorts and cleanliness dJ'ing cutting, drilling, grinding, etc., and control board protection from . falling objects. In addition, the operator has the authority to stop work at any. time. l Based on the above, it is concluded that the necessary measures have been taken to ensure and verify that the CN modifications and their possible malfunctions are analyzed by or enveloped byFSAR analyses. 3.(C) The margin Eof safety as defined in the Basis for any Technical Specification or in the FSAR has not been reduced. LCO 4.1.9_ was reviewed since it is applicable when the reactor is shutdown with core ' inlet orifice valves at any position and the reactor will be depressurized for the defueling process. These CN modifications will have no affect'on LCO 4.1.9 requirements. LCO 4.5.1 identifies the requirements for reactor _ building integrity and is applicable-during the defueling process. The CN modifications-have no deleterious effects on the Reactor Building Exhaust System. LCO' 4.5.2 is applicable during reactor vessel internal maintenance with irradiated fuel within the vessel which requires removal of both primary and secondary closures. The defueling process is internal maintenance. The modifications performed by CN 2939 are to expedite the process and to provide some of the components (primarily PDC-1156) necessary to maintain reactor pressure in accordance with LCO 4.5.2.a). LCOs 4.7.1, 4.7.2 and 4.7.3 address fuel handling, Fuel Handling Machine (FHM) and fuel storage. This CN enhances the ability to maintain the reactor depressurized (LCO 4.7.1.a), control pressure in the FHM (LCO 4.7.2.a), and maintain fuel storage well pressure at approximately atmospheric. These enhancements will help ensure no leakage of potentially contaminated gaseous effluents to the Reactor Building environs. f

7, 4 F i Page 2.6 lc o ELCOs 'and ESRs 8.1.1.and 8.1.2 are applicable to all releases of potentially radioactive effluents, and will ensure planned releases remain within 10CFR20 limits. This CN adequately ~ considered and ' designed new installations (with HP input) to comply with the ELCOs, ESRs and.10CFR20. l a l L f I I G j ) . I q I i; )

- _. _ ~ _ F RT ST. VRAIN NUCLEAR GENERATING STATION OPublic. $9fVICOs Pusuc sanVICE COMPANY OF COLORADO / OTHjR ENVittONMENTAL EVALUATION AfM PAGE 3 TYPE: % CN Overall O CN Submittal Other Are all measurable nonradiological effects of this activity confined to the on-site areas previously disturbed during site preparation, plant construction or previous plant operation? % Yes O No S MNU$t%2 4.TS& u) nVJ, 4-b s CAJ we__ State basis w C,% + d 1-o n Rm 4>,~ a., d /n 7 trnla Eu)IDnp. Is the activity required to achieve compliance with Federal, State or local environmental regulations? O Yes )( No Applicable Regulations NOTE: If eother answer is Yes, the activrty does not involve an unrewowed environmentaloverion. Sign and date the form. If both answers are No, the activory has tne potentral for crearong an unrewowed onwronmentalovestoon. Complete the remainder of this evaluation form. 1. Is the activity identified in the final environmental statement (FES) O Yes O No Or Supplementary Environmental Documents (See 0 3P Identify documents and document sections reviewed t 2. Determine whether or not the activity involved is an unreviewed environmental question using the following guidelines. (If the answer to any of the following questions is Yes, then this activity involves an unreviewed environmental question.) (A) Will this activity result in a significant increase in any adverse environmental impact previously evaluated in the FES? O Yes O No State basis I-l (B) Will this activity result in a significant change in the types, or a significant increase in the amounts of effluents, or a significant increase in the authorized power level? O Yes O No State basis L (C) Does this activity involve an environmental matter not previously reviewed and evaluated in the FES? O Yes O No State basis Does the activity involve an unreviewed environmental question? O Yes O No l0 4 7-8Y Approved E M G ro .inlJY~.M.. -{/n l ,1931-By ~ a -....n FORM tCl 372 24-4310 ( 0-33 l I,

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1 CN 2 BY ESN PAGr 8 M / FORM NO. 373 30 3650 P - &. I; markups ( AR1370SK1-6)' depict the system lir.sup suggested for PCRV. pressure control. Note that an alternate lineup (Ref. AR1370 'SK2A & 2B) through an H., getter unit into the pumpdown line is also possible. However, no' eredit for tritium removal can be taken because Getter sponge tempera'ture will be low. t The perception of limited flow capacity of the purge pump: discharge piping stems from observed opening of PCV 1316 L (.3 psig set pt.) recirculation valve .during purge vacuum pump operation. while evacuating the ATC. The perception of i limited flow capacity is supported by an informal. ATC pumpdown-test dated 3/3/89 (Ref. AR1370 Attachment C). This test evacuated the 400 cu. ft. ATC from .35" Hg. Vac. (12.13 palan to 19.9" Hg. vac. (2.553 psia) in the first 15 minutes and to 21.45" Hg. Vac. (1.794 psla) in an additional 10 minutes using C1302. This indicates a pumping rate during-the first 15 minutes of approximately 40 acfm 1.e. L pumping rate acfm equals: L Vol. ft." Press Initial 400 12.13 In = In = 41.56 time min. Press final 15 2.553 Since the recirculation valve PCV 1316 is verbally reported.to have l 1') been cycling during this test the pumping rate must also be recognized as purge pump discharge piping and/or' gas waste system flow. capacity. This limited flow capacity has the effect of significantly extending evacuation time of the ATC which was 25 minutes by test IRef. AR1370- Attachment 'C). A much improved pumpdown rate of 150 acfm is possible through routing purge pump L discharge directly to system 73 filters as was done during the 1989 PCRV' evacuation. During the 1989 PCRV evacuation using one' purge vacuum pump disenarging to Sy!. 72 PCRV pressure decreased from 5.72 l= psia-to 2.1. psia in 240 minutes. l 36000 ft.3 5.72 psia pump rate = X in = 150.3 a.fm 240 min. 2.1 psia l l O -A A-- w

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l ATC pumpdown time could be decreased to 5 minutes L through purge pump discharge directly to System 73. Using the same pressure parameters (P initial = 12.13-psia, P final = 2.553 psia during the ~ -first 15 minutes'of the 3/3/85 test).the new pumpdown time = ~400 12.13 X in = 4.15 minutes ITD M say approximately 5 minutes. A possible time saving of 20 minutes per .ATC purge operation is therefore indicated. Based on 16 purges per region the time saved could add up to 5 hours per, region. Any time saved on FHM avacuatign would be insignificant because of the small volume (4 ft. ) l-E involved. It 1s therefore proposed that capability for purge pump disenarge directly to System 73 be added. P&I marxups per AR1370 SF. 11-& 13 depict the proposed additicn. 1 u l l. I -a--

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Interoffice: Momoi N h h h M gt PPC-89-2519. j [ 'CN. E88k BY_ W M .DATE:.. August 3,: 1989 = L PAGE' 8 7 [.c l TO:' Hugh J. O'Hagan, Outage Manager LFROM: . Timothy E. Schleiger, Superintencent of Chemistry 2 W and Radiation Protection 1 i SUBJ: PRIMARY'C00LANT ACTIVITY AFTER FINAL REACTOR SHUTDOWN. FOR _ DEFUELING 4 The task fPrimary Coolant' Activity After Final' Reactor' Shutdown For. -Defueling".is complete. The intent of this task was to determine if the circulating activity. in the Primary' Coolant System.would be sufficiently reduced 30 days after final. reactor shutdown to preclude any further need for the use of the Helium Purification Regeneration System, i An analysis was performed utilizing current circulating activity at 1 85 reactor power'and worse case tritium activity. This analysis-shows. that continued use of the purification system for a period of 14 days following final-reactor-shutdown would reduce-circulating activity in the Primary Coolant System to 178.2 uC1 and 0.0021uti of tritium. Based :upon :this activity, dose calculations were performed in. .accordance with the'FSV ODCM and demonstrated that there would be no measurable exposure to a member of the general public should the need e arise to perform'a controlled release of the PCRV. inventory directly to the atmosphere. .a The-benefit to be derived from utilization of the " Regeneration Pit"- during defueling from a cost'- saving standpoint compared to the~ -potential; : insignificant exposure that could be received by a member - f the general public is justified. o It is imperative that if this option is implemented, that clear procedural requirements must be established to minimize.and control 4 releases from the PCRV to the maximum extent practicable in order to ensure compliance with ALARA considerations. t 4 t'

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? N Intoroffice Memo l: PPC-89-2519- .%hh Mg*[ -August 3. 1989-Page 2 l 4-ll M*W*" CN 2Y3$ BY 8 /J5f PAGT 8 A /' Should you have any questions, I can be contacted at extension ' FORM too. 373 30 3660 - y e M l l. 1 Timothy E. SchleigeV Superintendent of. Chemistry and c, Radiation Protection TES/bhb-t o ? l~ t 'r D 'i 'd ( ^ l .i !1 l o 1: [ p oI- - FORM (C) 31422416 f .w- " ' " '}}

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