ML20148P402

Rev 5 to Vynp Process Control Program

ML20148P402
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/23/1996
From:	Houle M, Wanczyk R VERMONT YANKEE NUCLEAR POWER CORP.
To:
Shared Package
ML20148P392	List: ML20148P395 ML20148P402 ML20148P408
References
PROC-960423, NUDOCS 9707020220
Download: ML20148P402 (24)

Text

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VERMONT YANKEE NUCLEAR POWER CORPORATION PROCESS CONTROL PROGRAM REV 5 4/23/96 1

Submitted A4/'P Ra at*p Pro /ction Manager Approved W W. M Y2/fr. 9/pdN'7 PORC 9 Approved A /

t Manager [//

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i APPENDIX A  ;

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

. . Supplemental Information for 1996 I I

Facility: Vermont Yankee Nuclear Power Station )

l Licensee: Vermont Yankee Nuclear Power Corooration 1A. TECHNICAL SPECIFICATION LIMITS - DOSE AND DOSE RATE Technical Specification and Category Limit

a. Noble Gases 3.8.E.1 Total body dose rate 500 mrem /yr 3.8.E.1 Skin dose rate 3000. mrem /yr 3.8.F.1 Gamma air dose 5 mrad in a quarter 3.8.F.1 Gamma air dose 10 mrad in a year 3.8.F.1 Beta air dose 10 mrad in a quarter 3.8.F.1 Beta air dose 20 mrad in a year .

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b. Iodine-131. Iodine-133. Tritium and Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days 1

3.8.E.1 Organ dose rate 1500 mrem /yr 3.8.G.1 Organ dose 7.5 mrem in a quarter 3.8.G.1 Organ dose 15 mrem in a year

c. Liauids 3.8.B.1 Total body dose 1.5 mrem in a quarter 3.8.B.1 Total body dose 3 mrem in a year 3.8.B.1 Organ dose 5 mrem in a quarter 3.8.8.1 Organ dose 10 mrem in a year i

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f APPENDIX A (Continued)

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

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TECHNICAL SPECIFICATION LIMITS - CONCENTRATION 2A.

Technical Specification 'and Cateoorv Limit 1

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a. Noble Gases No MPC Limits (No ECL Limits) b'. Iodine-131. Iodine-133. Tritium and Radionuclides in Particulate Form .

With Half-Lives Greater Than 8 Days No MPC Limits (No ECL Limits) c Liauids 3.8.A.1 Total fraction of MPC (ECL) excluding noble gases (10CFR20. Appendix B.

Table II, Column 2): 11.0 l

3.8.A.1 Total noble gas concentration: 12E-04 pC1/cc

3. AVERAGE ENERGY Provided below are the average energy (E) of the radionuclide mixture in releases of fission and activation gases, if applicable. I

a. Average gamma energy: Not Applicable

b. Average beta energy: Not Applicable

4. MEASUREMENTS AND APPR0XIMATIONS OF TOTAL RADI0 ACTIVITY Provided below are the methods used to measure or approximate the total radioactivity in effluents and the methods used to determine cadionuclide composition.

a. Fission and Activation Gases Continuous stack monitors monitor the gross Noble Gas radioactivity released from the plant stack. Because release rates are normally below the detection limit of these monitors, periodic grab samples anao A-2

1 APPENDIX A (Continued)

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT are taken and analyzed for the gaseous isotopes present. These are used to calculate the individual isotopic releases indicated in Table 18 and the totals of Table 1A. The error involved in these steps may be approximately 123 percent. .

b. Iodines Continuous isokinetic samples are drawn from the plant stack through a particulate filter and charcoal cartridge. The filters and cartridges are normally removed weekly and are analyzed for Iodine-131, 132. 133, 134, and 135. The error involved in these steps may oc approximately 118 percent.

c. Particulates The particulate filters described in b. above are also counted for particulate radioactivity. The error involved in this remple is also approximately 118 percent.

d. Tritium Grab samples from the plant stack are taken monthly through a cold trap collection device and analyzed for tritium. The error involved in this sample is approximately il5 percent.

e. Waste Oil Prior to issuing the permit to burn a drum of radioactively contaminated waste oil, one liter of the oil is analyzed by gamma spectroscopy to determine concentrations of radionuclides that meet or exceed the LLD for all of the liquid phase radionuclides listed in Technical Specification Table 4.8.1. Samples that have a visible water layer ara not analyzed. The water must first be removed from the drum of oil and resampled.

Monthly, samples from drums that were issued burn permits are sent to the E-Lab for compositing and analysis. The E-Lab analyzes for tritium, alpha, Fe-55 Sr-89, and Sr-90 on the composite sample.

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. APPENDIX A (Continued) ,

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT

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i The waste oil ' samples are liquid effluents that end up-as a gaseous i . ground level release.

Liauid Effluents

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3 Radioactive. liquid effluents released'from the facility are 'l-continuously monitored. Measurements are also made on a l representative sample of each batch of radioactive liquid effluents l released. For each' batch, station records are retained of the total j activity (mC1) released, concentration (pC1/ml) of gross l

radioactivity, volume (liters), and approximate total quantity of water (liters) te ed to dilute the liquid effluent prior to release to the Connectiu. River.

Each batch of radioactive liquid effluent releases is analyzed for gross gamma and. gamma isotopic radioactivity. A monthly l proportional compos.ite sample, comprising an aliquot of each batch l released during a month, is analyzed for tritium and gross alpha .j radioactivity. A quarterly proportional composite sample, .j comprising an aliquot of each batch released during a quarter, is  ;

analyzed for Sr-89, Sr-90, and Fe-55. q t

5. BATCH RELEASES

a. Liouid There were no routine liquid batch releases during the reporting y.riod.

b. Gaseous There were no routine gaseous batch releases during the reporting period.

6. ABNORMAL RELEASES

a. Liauid There were no nonroutine liquid releases during the reporting period, am6o A4

APPENDIX A (Continued)

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT l

b. Gaseous There were no nonroutine gaseous releases during the reporting period. I l

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APPENDIX B

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l LIQUID HOLOUP TANKS i a

Reauirement: Technical Specification 3.8.D.1 limits the quantity of radioactive material contained in any outside tank. With the quantity of radioactive material in any outside tank exceeding- "

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.the limits of' Technical Specification 3.8.0.1, a descript, ion of

'the events leading to this condition is required in.the next Annual Effluent Release Report per Technical Specification  !

6.7.C.1.

Response: The limits of Technical Specification 3.8.D.1 were not exceeded during this reporting period. ,

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APPENDIX C~

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RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSiCUMENTATION Reautrement: Radioactive liquid effluent monitoring instrumentation channels are required to be operable in accordance with Technical )

Specification Table 3.9.1. If an inoperable radioactive ' liquid effluent monitoring instrument is not returned to operabl,e status prior to a release pursuant to Note 4 of Table 3.9.1, an explanation in the next Annual Effluent Release Report of the reason (s) for delay in correcting.the inoperability are required l

.per Technical Specification 6.7.C.1.

Response: Since the requirements of Technical Specification Table 3.9.1 governing the operability of radioactive liquid effluent l

hionitoring instrumentation were met for this reporting period. no  !

response is required, j I

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l l-APPENDIX D RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION f Reautrement: Radioactive gaseous effluent monitoring instrumentation channels are required to be operable in accordance with Techrical Specification Table 3.9.2. If inoperable gaseous effluent monitoring instrumentation is not returned to operable status within 30 days pursuant to Note 5 of Table 3.9.2, an explanation in the.next Annual Effluent Release Report of the reason (s) for the delay in correcting the'inoperability is required per Technical Specification 6.7.C.1.

1 Response: Since the requirements of Technical Specification Table 3.9.2 governing the operability of radioactive gaseous effluent monitoring instrumentation were met for this reporting period, no response is required I

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1 APPENDIX E RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM l

Reautrement: The radiological environmental monitoring program is conducted in accordance with Technical Specification 3.9.C. With milk samples no longer available from one or more of the sample locations  ;

required by Technical Specification Table 3.9.3, Technica,1 l Specification 6.7.C.1 requires the following to be included in the next Annual Effluent Release Report: '(1) identify the

! cause(s) of the sample (s) no longer being available, (2) identify the new location (s) for obtaining available replacement samples and (3) include revised ODCM figure (s) and table (s') reflecting .

the new location (s). )

l Response: The Back Tracks Farm in Vernon, Vermont went out of business on l l '

L April 9, 1996. As a consequence, Table 4.1 Parts 3.a and 3.c, as well as Figure 4-2 of the ODCM, have been revtsed to remove this milk and silage sampling location from the Radiological Environmental Monitoring Program (REMP). No new. replacement stations need to be added to the program since the required l

number of highest-dose-potential farms, as listed in the 1995 l l Land Use Census, are already part of the REMP. The revised ODCM l, figure and table are included in the attachment to Appendix H. l 1

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! I l APPENDIX F l

l LAND USE CENSUS l Reauirement: A land use census is conducted in accordance with Technical Specification 3.9.D. With a land use census identifying a location (s) which yields at least a 20 percent greater dose or l dose commitment than the values currently being calculate,d in I Techni .; 5pecification 4.8.G,1, Technical Specification 6.7.C.1 requires the identification of the new location (s) in the next Annual Effluent Release Report.

l Response: The Land Use Census was completed in the third quarter of 1996. 3 No locations yielded a 20 percent greater dose or dose commitment I than the values currently being calculated in Technical Specification 4.8.G.1.

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APPENDIX G I

L PROCESS CONTROL PROGRAM Reauirement: Technical Specification 6.12.A.1_ requires that' licensee initiated L changes to the Process Control Program (PCP) be submitted to the l Commission in the Annual Radioactive Effluent Release Report for 1 the period in which the change (s) was made. ,

j Response: The following changes were made to the Process Control Program (PCP) and issued as Revision 5 during this reporting period.

Title Page: Changed approval by Operations Manager to VP/M00 as stated in Tech Specs.

I Section 1: Added reference to TS 4.8.N; added statement to review and approve vendor's PCP program by PORC and VP/M00.

Section 2: Added statement to air dry filters 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or as determined by RWS. Also added method to dispose of filters above dose limitations per 49 CFR.

Section 3: Added reference to OP 2153. Remove reference to remote dewatering which is no longer used.

Section 4: Removed reference concerning the use of a vacuum pump to dewater the liners. This method is no longer used.

Section 5: Added reference to OP-2512.

i Section 6: None. )

l Section 7: Added reference to MSDS via AP-0620. ]

1 Section 8: Same as Section 7.

Section 9: Same as Section 7.

Section 10: None, t

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l APPENDIX G I (Continued) I PROCESS CONTROL PROGRAM Section 11: Added reference to the use of an approved outside lab.

Section 12: Added this new section - Mixed Waste. .

Page 4 was revised to incorporate procedure title and revision changes. OP-2511 and OP-2512.

This revision does not affect Technical Specifications and does not affect any system or processes described in the FSAR.

The revised Process Control Program is attached.

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1 VERMONT YANKEE NUCLEAR POWER CORPORATION 1 PROCESS CONTROL PROGRAM REV 5 4/23/96 l 1

Submitted /H/P i Ra at*fn Pro /ction Manager l Approved W W. M V7/fr. 9/o-MN'7 PORC 9 Approved e i I t Manager [//

Approved N l l VP, MOO l i

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i VERMONT YANKEE NUCLEAR POWER CORPORATION ,

PROCESS CONTROL PROGRAM

Introduction:

'.~he Vermont Yankee Nuclear Power Corporation Process Control Program (PCP) describes the administrative and technical controls on the radioactive waste systems which provide assurance that Vermont Yankee meets federal shipping and burial site requirements.

The PCP complies with Technical Specification 6.12 by describing process parameters, controls, tests, sampling and analysis to ensure compliance with 10 CFR 20,10 CFR 71, and 10 CFR 61 Energy,49 CFR 172-173 Transportation, state, and burial site regulation requirements.

1.0 Solidification Vermont Yankee Nuclear Power Corporation does not routinely solidify liquid waste.

If the use of solidification to dispose of may liquid waste is required,it will be done by l an outside vendor under the vendor's PCP. The vendor PCP will be reviewed and l a,pproved by the Plant Health Physicist, the Radiation Protection Manager, PORC and' l VP, MOO prior to implementation. This review is to identify that there is sufficient supporting documentation of the vendor's PCP to give assurance that the final product will meet all requirements for transport and burial, and that sufficient procedural l controls exist to assure safe operations. [TS 4.8.N]

2.0 Cartridae Filter Elements l Low activity cartridge filter elements will be air dried (~ 24hr or as' determined by the l Radwaste Supervisor) and handled as dry active waste. Filters determined to be above l the dose limitations per 49CFR, will be placed in casks and shipped for disposal.

3.0 Resins I Normal operations produce radioactive waste in the form of depleted resins. These resins are processed in the burial container using a rapid dewatering systern !RDS-l 1000) manufactured by Chem-Nuclear Systems, Inc. [OP 2153]-

The system has been tested, by Chem-Nuclear, for certification in meeting the Bamwell Site Criteria and disposal requirement for free standing liquid. These tests are described in Chem-Nuclear's Topical Report on the RDS-1000 Radioactive Waste Dewatering System. In addition, to comply with the statement, "Any liquids present ir. waste packages shall .be non-corrosive with respect to the container," Vermont Yankee tested the pH of various resin mixtures used by the plant in solution with water. The range was found to be 4.2 - 8.4. A solution is not considered corrosive if the pH is greater than 4.0.

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i 1 A resin sample is taken from each' liner prior to shipment. The sample is counted to '

determine the activity and waste classification. Class A resins that exceed 1.0 yCi/cc of isotopes with greater than 5 year half-lives and all Class B and C resins will be s disposed of in an approved High integrity Container (HIC). l l

Vendor supplied or temporary methods of processing resins may be used in lieu of the above process provided that the vendor or temporary process meets the requirements of quality described above and does not conflict with accepted burial criteria or safety

! requirements. Such methods will be reviewed and approved by the Plant Health  !

Physicist and the Radiation Protection Manager prior to implementation.

4.0 Filter Liners During refueling outages and normal operation, liquid radwaste processing may require I use of a decanting filter on the condensate phase separators. A floating suction is used to decant the water and resin into a filter liner. Filtered water is pumped from the l liner. Th'e liner is dewatered in accordance with OP 2511 (MOOID9409 03) such that l the burial site criterion for free-standing water is met. A resin sample is taken from the j l liner and analyzed to determine the activity and waste classification.

f 5.0 Drv Active Waste (DAVI) l DAW is compacted, as practical, or shipped to a vendor that sorts the material for 1 processing or recycling. All DAW is examined before being compacted or shipped. l Any liquids or items found that would compromiso the integrity of the package ~are l removed and separated as specified by procedure. [OP 25121 Containers used for DAW shipments meet the criteria of 49 CFR 173.425a. or b. "No leakage of

• radioactive material," as specified in 49 CFR 173.425.b.1 will be. met provided that no radioactive materials in quantities equal to or exceeding those specified in 49 CFR

,173.443 are detected on the external surfaces of the package at any time during shipment.

6.0 Chelatino Aaents in order to comply with 10 CFR 20 Appendix F, chelating agents are controlled by the plant chemistry department using procedure AP 0620.

7.0 Exolosive Waste No waste capable of detonation or of explosive decomposition or reaction will be l disposed as per 10 CFR 61.56(a)(4). Refer to MSDS via AP 0620.

8.0 Toxic Waste

! No waste capable of generating toxic gases, vapors, or fumes will be disposed as per

'j 10 CFR 61.56(a)(5). Refer to MSDS via AP 0620.

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4 9.0 Pyroohoric Waste j No waste that is pyrophoric will be disposed as per 10 CFR 61.56(aliG). Refer to l MSDS via AP 0620.

10.0 Hioh Inteority Containers (HICs)

Vermont Yankee Nuclear Power Corporation has contracted with various suppiiers of i approved HICs. South Carolina has approved PCPs for HICs used by Vermont Yankee. .

Any HIC Vermont Yankee may choose to use at some future time, will meet all applicable requirements.

11.0 Waste Class Determination l Along with an approved outside laboratory, Vermont Yankee periodically performs laboratory analysis on all waste streams to determine the activity of radionuclides l listed in Tables 1 and 2 of 10 CFR 61. Correlation analysis verifies that the relative concentration of each radionuclide, with respect to the overall activity in a given Vermont Yankee waste stream, remains constant over time. A set of scaling factors is determined which allows the activity of 10 CFR 61 radionuclides to be estimated using the results of gamma spectrometric analysis or direct gamma dose rate measurements.

For' resin wastes, analysis is performed on samples of each source of resin comprising the contents of a burial container, Scaling factors are applied to the activity of radionuclides identified by gamma spectrometry analysis to determine the activity of those radionuclides which are not detected in the gamma spectrum. ,

For DAW, dose rate-to-curie conversion calculations are performed to determine the total activity present in a container. Scaling factors are applied to the container's total curie content to determine the activity of individual radionuclides. -

Specific procedures for determining 10 CFR 61 scaling factors are contained'in OP

. 2527, " Sampling and Analysis for Radwaste Classification." Once the activity of each radionuclide in a burial container is estimated, the waste classification is derived using 2

methods required by 10 CFR 61. Specific procedures for waste class determination

. are contained in AP 0504, " Shipment of Radioactive Material."

l 12.0 Mixed Waste

l. No mixed waste will be disposed as per 10 CFR 61.56(a)(8) unless properly treated.

l Refer to MSDS via AP 0620.

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PROCEDURES WHICH IMPLEMENT THE PCP  ;

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1. AP 0504 Shipment of Radioactive Materials l 2. OP 2511 Radwaste Cask / Liner Handling

3. OP 2527 Sampling and Analysis for Radwaste Classification

4. OF 2151 Liquid Radwaste

5. OP 2153 Solid Radwaste

6. AP 0620 Chemical Material Use l 7. OP 2512 Radwaste Drum, Box and Sealand Handling ,

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APPENDIX H  !

0FF-SITE DOSE CALCULATION MANUAL .

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j Reauirement: Technical Specification 6.13. A.1 requires that licensee initiated changes to the Off-Site Dose Calculation Manual (0DCM) be i L submitted to the Commission in the Annual Radioactive Effluent l Release Report for the period in which'the change (s) was ,made effective.  ;

Response: Revision 20 to the Off-Site Dose Calculation Manual was made

• during this reporting period.

The major changes included in Revision 20 to the ODCM are:

(1) Page 3-51: Direct off-site dose due to N-16 decay from tne turbine is calculated using Equation 3-27 in the ODCH. One variable in this equation is the gross electric output during periods of interest. In 1995, new turbine rotors-were installed which.had the effect of increasing the electric output of the plant due to higher turbine efficiency. However, the actual mechanism upon which the dose from N-16 is derived, involved the carryover of N-16  ;

in the steam flow from the reactor through the turbine, and not directly the electric output of the plant. The measurements that related N-16 dose with power level used  !

gross electric output as a readily available parameter to represent the relative amount of steam that passed through the turbine. With the increase in efficiency due to the new turbine rotors, the correlation between steam flow and electric output has changed slightly (i.e., there is more gross electric output for the same steam flow, and therefore the same N-16 carryover as before). The original turbine generator heat balance (FSAR Figure 1.6-2) indicated'a gross heat rate of 10,123 Btu / kwhr. The revised calculated heat balance-(General Electric Co.

Drawing 510HB362 Rev. 2) shows a gross heat rate of 9891 Btu / kwhr, or approximately 2.3% increase in efficiency. As a consequence, a conservative 2% reduction in the N-16 dose conversion factor of ODCM Equation 3-27 has been made to adjust the expected dose as a function of new gross electric output of the turbine.

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APPENDIX H (Continued) 0FF-SITE DOSE CALCULATION MANUAL (2) Page 3-5: The old turbine rotors (and casings), as noted above, were placed in storage sheds near the west site boundary. Surveys of the rotors and casings indicate low level surface contamination (principally Co-60). Opse rate projections from these components based on the surveys taken at the storage location indicate that the annual impact on the maximum west site boundary from direct dose is about 0.2 mrem. Since the turbine hall and other fixed sources already account for direct radiation doses on the order of about 15 mrem / year (reported to the nearest tenth of a mrem) dose contributions of less than 0.1 mrem / year are not considered significant with respect to their impact on the total dose. A commitment is added to the ODCM to include in the annual estimate of total direct dose from all sources the projected site boundary dose from the turbine rotor storage sheds as long as the rotors are in storage and the projected dose from them is greater than 0.1 mrem / year.

i (3) Table 1.1-12: Isotopic analysis of recent gas samples indicates, that at times. Co-57 is detected in effluent releases to the environment. This radionuclide was not included in the original listing of dose factors for Method I dose calculations since it was not included in the data library listings of Regulatory Guide 1.109 which formed the bases of all 00CM dose factors. In response to the identification of Co-57 in gas effluents, new dose factors have been developed using the same modeling approach and data bases as used by the NRC in the derivation of the original set of Regulatory Guide 1.109 dose factors. The resulting dose factors are, therefore, consistent with the methods already accepted and used in Regulatory Guide 1.109 and the ODCM. In addition, new dose factors for two other radionuclides (Se-75 and Sn-113),

that were not included in the regulatory guide listing, have been developed in the same way as for Co-57. This is because other plants have reported detecting these radionuclides in effluents. Table 1.1-12 dose factors have been recalculated with the expanded library to add these three radionuclides to the ODCM listing.

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1-APPENDIX H

.(Continued) l OFF-SITE DOSE CALCULATION MANUAL It should also be noted that the recalculation of I-133 resulted in a slight difference in the last decimal place of the original dose factor listed in the ODCM. This is believed to be the result of a slight change in the last digit round off due to the use of an updated half-life value for this radionuclide. No significant differences in off-site doses would result from this change, but it should be reflected in plant procedures to ensure consistency.with the ODCM.

-(4) Page vi, Appendix E: By letters, dated November 18, 1991 and July 10, 1992 Vermont Yankee submitted to the NRC a request for approval of proposed procedures for in-place disposal of slightly contaminated radioactive soil under the Chem Lab floor which had resulted from a break in a chem sink drain line. The NRC responded to this request with their_ approval (letter NVY 96-48, dated March 7, 1996) documented in a Safety Evaluation. The NRC acceptance of our plans to dispose of the soil in place required that their SE be incorporated into the ODCM as an Appendix. A new Appendix E containing the NRC SE has been added to the ODCM to complete this requirement (LAI No.11308).

(5) Pages 4-2, 4-2a, and 4-5: On April 9,1996, it was identified that the Back Tracks Farm in Vernon, Vermont (location TM-10 and TC-10) had gone out of business. It has been added to the Radiological Environmental Monitoring Program (REMP) in the fall of 1995 as a result of the 1995 Land Use Census ranking it among the highest potential receptor locations that warranted inclusion in the monitoring program. As a consequence, Table 4.1 Parts 3.a and 3.c. as well as Figure 4-2, have been revised to remove this milk and silage sampling location from the REMP. No new replacement stations need to be added to the program since the required number of highest-dose-potential farms, as listed in the 1995 Land Use Census, are already part of the REMP.

l (6) Page 5-17: Current plant calibration procedu:es use Xe-133 as the reference radionuclide to determine counting

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APPENDIX H (Continued) 0FF-SITE DOSE CALCULATION MANUAL efficiencies of the Stack Gas I and II, and A0G monitors.

This is based on the combination of its expected presence in the off-gas, and a relatively long half-life that make:

it practical to detect before decay eliminates it. .Xe-133 is a commonly used gas in the industry for monitor calibrations. During normal plant operations where off-gas is processed through the A0G, short-lived gases are decayed away before being seen by the monitors, leaving only long-lived Xe-133 and Kr-85. The use of Xe-133 instead of Kr-85, or a mix of the two, is based on the fact that the expected response'of the detectors is conservative for Xe-133 in comparison to Kr-85. If short-lived noble gases were present in significant amounts due to some off-normal condition, the use of Xe-133 as a reference gas for counting efficiercy would still lead to a conservative monitor response relative to Xe-133 since short-lived gas mixes tend.to have higher response factors (i.e., high energy noble gases will cause the monitor to respond as if a higher concentration of radioactivity were being discharged than actually is). The new wording also clarifies that grab samples and lab analysis of gas releases can be used to determine the release rate of noble l gases from the stack or A0G at any time. This ensures operational flexibility if menitor readings are suspected j of over responding to actual release conditions. This is J considered an administrative change to simplify plant procedures.

l (7) Page 6-10: Figure 6-2 illustrates the basic operational i layout of the Advanced Off-Gas (A0G) System. Based on the )

results of an A0G System review, Figure 6-2 has been corrected to indicate that the A0G bypass line that provides a discharge path around the moisture removal / dryer subsystem and charcoal adsorber subsystem feeds back into l the main discharge line downstream of the A0G noble gas I monitors (RAN-0G-3127 and RAN-0G-3128) and before the final j 60% section of the original delay pipe. This had been incorrectly shown that the bypass line returned gas flow I

upstream of these two monitors. This is considered an l administrative change to reflect 'as built" conditions. ,

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APPENDIX H (Continued) 0FF-SITE DOSE CALCULATION MANUAL (8) Page 4-1: In 1995, the EPA announced that they would not be able to supply a full range of environmental control samples in 1996 for use in laboratory intercomparison studies program in which all nuclear power plants are required to participate. Vermont Technical Specification 3/4.9.E requires that the approved NRC Intercomparison Program used by the laboratory performing environmental analyses be identified in the ODCM. Section 4.1 of the ODCM has been updated to indicate the various NIST traceable quality assurance programs in which the YAEL participates. A recent NRC Region I inspection of the YAEL 0A Programs concluded that the intercomparison studies in which the YAEL participates satisfy NRC requirements to meet the Technical Specification commitments.

(9) Pages 3-14 and 3-20: A clarification has been included in.

the definition of stack release quantities that provides for direct isotopic measurement (grab samples with laboratory analysis) of the radionuclide mixes in the gas stream being discharged from the stack, and the alternate method of inferring the radionuclide mixes at the stack by using SJAE measurements. These two approaches are part of current plant procedures, and provides operational flexibility to estimate the continuous radionuclide release rate from the plant stack. This is considered an administrative change to reflect existing practices.

(10) Pages 5-13, 5-14, 5-15, and Appendix A: With the routing of the turbine hall exhaust air to the plant stack, and the replacement of the stack gas monitors (I and II) with new detectors, both effluent flow rate and monitor response factors have changed slightly from values originally used in example calculations used to illustrate the use of various dose and dose rate applications. This revision updates these illustrations to provide plant staff with l

! examples that reflect information more relevant to the current plant configuration. These changes are considered administrative in nature in that no change is made in the methodology used to calculate doses.

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' APPENDIX H l p (Continued)

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l OFF-SITE DOSE CALCULATION MANUAL -

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The above changes will not t reduce the accuracy or-reliability of

-the dose calculations or setpoint determinations previously

. approved for use in the ODCM. 'This. conclusion is based on the nature'of'the changes that: (a) adjust existing methodology for calculating direct dose by factoring changes in turbine >

efficiency to maintain the-same level of N 16 impact as a l function of electric power as previously used (b)' include  ;

additional fixed radiation sources in estimating a better total i site boundary dose. (c) use the existing dose methodology to  !

expand the listing of radionuclides that can be. considered in calculating Method I dose impacts. (d)' expand the description of l gas monitor calibrations to reference Xe-133:as the standard-

)

reference gas in keeping with common industry practices, (e) ]

clarify existing plant procedures for direct measurement of J effluent ga's mixes in addition- to' estimates derived' from SJAE. 3 measurement,~and (f) describe administrative changes that update-system flow diagrams to existing plant configurations, add new ,

documentation for the approval to dispose of slightly.  !

contaminated soil, replace REMP sampling locations due to sample-unavailability, describe the currently approved intercomparison 0A Program at the YAEL, and show examples of the type of calculations that can be addressed by'the ODCM methods.

The revised pages from Revision 20.to the ODCM are attached.

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