ML20024G396

Forwards Draft Interim Tech Specs Section 2.4 Re Radioactive Effluents

ML20024G396
Person / Time
Site:	Monticello
Issue date:	03/01/1976
From:	Mayer L NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9102110422
Download: ML20024G396 (41)

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March 1, 1976 ,,.

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N M-4 l fV  ?/n {0, Mr. Victor Stello, Director

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Dear Mr. Stello:

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N ,,A MONTICELLO NUCLEAR GENERATINI ' 'NT * --

DOCKET NO. 50-263 LICENSE NO. V 4-22 Draf t Interim Technical Specifications Radioactive Effluents Our letter of June 13, 1975 furnished copies of a draf t Section 2.4 covering radioactive effluents. Wese effluent Technical Specifications would be issued on an interim basis until new specifications based on the " Adopted Appendix 1" become ef fective.

Following your review of our draf t Section 2.4, members of your staff communicated a number of comments to us and equerted that revisions be made in the draf t to reficct those comments. Accordingly, attached are forty copies of a revised draft which we believe is responsive to the comments and consistent with the comunications we have had with your staff. For convenience, rather than revising affected pages, this new draf t completely supersedes the draft furnished with our letter of June 13.

Certain portions of these interim Technical Specifications are gernaine to several issues in the ongoing Monticello Public Hearing. One of the intervenors in this hearing, the Minnesota Pollution Control Agency (MPCA), expressed a desire to participate in the development of interim Technical Specifications. Accordingly, a telephone ccnference call was held involving the NRC legal and technical staf f, and the MPCA.

At that time MPCA suggested several changes to be incorporated in the interim Technical Specifications and the draf t we are submitting has also been revised to include the MPCA suggestions.

When these interim Section 2.4 Technical Specifications are issued, cer-tain portions of the current Appendix A Technical Specification must also be issued in revised form. We appropriate revisions to the Appendix A Technical Specifications are described in Attachment A to thiP letter.

These revisions, insofar as contents, nomenclature and pagination are concerned, will have to be coordinated with the changes requested in our October 15, 1975 Submittal on the Radiation Environmental Monitoring Program.

9102110422 760301 2082 PDR ADOCK 05000263 p PDR

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NORTHERN OTATED POWER COMPANY

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f ,, Mr. Victor Stello March 1, 1976 Our letter of June 13, 1975 asked for a delayed implementation date for the Interim Technical Specifications on Radioactive Effluents. That specific request is no longer applicable because the Augmented Offgas System has been in continuous service following the modifications that were made during the Autumn 1975 refueling outage. In order to simplify release calculations, re-porting and compliance requirements, it is requested that these Interim Technical Specifications become effective on the first day of a calendar quarter. The NRC letter issuing the Interim Technical Specifications should specifically reference such an implementation schedule.

Yours very truly, b A-L. O. Mayer, P Manager, Nuclear Support Services LQ4/ECW/ deb cc: J. G. Keppler G. Charnoff MPCA Attn: J. W. Ferman attachment

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ATTACINENT A (NSP-NRC letter Deted March 1, 1976)

List of Monticello Appendix A Technical Specifications that Must be Deleted or Changed Upon Issuance of TS.B.2.4. Radioactive Ef fluents

1. Table of Cori . ;;s_t age iv Proposed Ch. 2c Revise the Table of Contents to properly reflect the deletions and changes made below. Particularly, SectL on 3.8 and 4.8 should be re-designated as Environmantal Monitoring Program, page 168; the "3.8 and 4.8 Bases" entry should reference page 172.

Reason for Change To correct the Table of Contents.

2. 3.2 Bases, page 68 Proposed Change In the last sentence of the second paragraph on page 68, change the words ". . .. maximum stack release rate limit allowed by specification

3. 8. A.1 is no t exceeded. " to read es follows: ". . . . max mum s tack release rate limit is not exceeded."

Reason for Change Since this section of TS.3.8 will be deleted, the specificatt on reference in the 3.2 basis is inappropriate,

3. Specification 3.8. A-E/4.8 A-F, pages le8-173A, Table 4.8.1 and Figur e 4.8.1. pages 174-176A Proposed Change Delete Technical Specifications 3.8. A-E and 4.8. A-E. Renumber TS.4.8.F as TS.4.8, Environmental Monitoring Program and move it to new page 168.

Move Table 4.8.1 to pages 169-171. Delete Figure 4.8.1.

Reason for Change The radioactive effluent technical specifications are now included in TS.B.2.4.

ATTACIBLENT A

4. Bases 3. 8. A-E/4.8. A-F. pages 177-179B Proposed Change Delete Bases 3.8.A-E and 4.8.A-E. Renumber Basis 4.8.F as 4.8 and move it to new page 172. On bottom of page 172, make a note "Next page is page 180".

Reason for Change the bases for the radioactive ef fluent technical specifications are now included in TS.B.2.4.

NOTE:

Revised Technical Specification pages showing these changes are included as pages 3-9 of this Attachment A.

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i 3.7 and 4.7 C;ntainment Systems 139 Primary Containment 139 A.

Standby Gas Treatment System 148 B.

Secondary Contaimment 150 C.

Drimary Containment Isolation Valves 151 D.

156 3.7 Bases 161 4.7 Bases h 168 di h}

3.8 and 4.8 Environmental Monitoring Program s:

9 172 -i 3.8 and 4.8 Bases >

3.9 and 4.9 Auxiliary Electrical Systems 180 Operational Requirements for Startup 180 A.

Operational Requirements for Continued Operation 181 B.

181

1. Offsite Power (Line) 182

2. Offsite Power (Transformers) iv REV

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Bases Continued:

3.2 For ef fective emergency core cooling for the small pipe break the HPCI or Automatic Pressure Relief system must function since for these breaks, reactor pressure does not decrease rapidly enough to allow either core spray nr LPCI to operate in time. The arrangement of the tripping contacts is such as to previde this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria is met. Reference Section 6.2.4 and 6.2.6 FSAR. The specification preserves the effectiveness of the system during periods of main-tenance, t es t in g , or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.

Two air ejector of f-gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector off-gas line. Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip or two downscale. There is a 30-minute delay before recombiner train inlet valve closure when the recombiners are in use and a 15-minut e delay before of f-gas isolation valve closure when the recombiners are bypassed in whichThe thetrip settings D reactor operator may take corrective action. Both instruments are required for trip. $

of the instruments are set so that the maximum stack release rate limit is not exceeded. &O e T 9

s Four radiation monitors are provided which initiate isolation of the reactor building and operation 3 of the standby gas treatment system. The monitors are located in the reactor building ventilation plenum and on the refueling floor. Any one upscale trip will cause the desired action. Trip settings of 3 mR/hr for the monitors in the ventilation duct are based upon initiating normal ventilation iso-lation and Standby Gas Treatment System operation so as not to exceed the maximum release rate limit allowed by Specification 3.8.A.1 for the reactor building vent. Trip settings of 100 mR/hr for the monitors on the refueling floor are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.

Although the operator will set the set points within the trip settings specified in Tables 3.2.1, 3.2.2, 3.2.3, and 3.2.4, the actual values of the various set points can dif fer appreciably f rom the value the operator is attempting to set. The deviations could be caused by inherent instrument e rror, j

operator setting error, drift of the set point, etc. Therefore, these deviations have been accounted for in the various transient analyses and the actual trip settings may vary by the following amounts.

68 3.2 BASES REV

re 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 4.8 Environmental Monitering Program The environmental monitoring program given in Table 4.8.1 shall be conducted. +%

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Table 4.8.1 I

SAMPLE COLLECTION AND ANALYSIS MONTICELLO NUCLEAR PLANT - ENVIRONMENTAL MONITORING PROGPAM Type of Sample Type of Analysis Collection Site Collection Frequency River Water GB, GS Upstream within 1000 ft of 1,itake canal, Weekly 311 (M) Downstream within 1000 ft of discharge canal.

St. Paul rav vater intake.

lake Water GB, GS, 31I 2 local lakes. Quarterly (water and >

ice conditions permitting Well Water CB, CS, 11 6 sites within 5 miles of plant site Quarterly ah including the Manticello well. 'g Precipitation CB, GS, 311, Meteorological Station at plant. Monthly 1311, 90Sr State llealth Dept Bldg - Mpls.

Lake and River GB, GS 2 local lakes. Semi-annually Bottom Sediment 137Cs Upstream of plant.

Downstream of plant.

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Plankton or Algae 2 local lakes. Semi-annually GB, GS,90 or Insects 137 Cs, Sr Upstream of plant. (when available)

Downstream of plant.

Aquatic GB GS, 2 local lakes. Quarterly vegetation 13 Cs Upstream of plant. (when available)

Downstream of plant. ,

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Table 4.8.1 Continued Type of Analysis Collection Site Collection Frequency f Type of Sample Upstream of plant. Quarterly  !

Clams GB, GS Downstream of plant. (when available)

Fish GB, GS Upstream of plant. Quarterly (water and Downstream of plant. ice conditions permitting)

Milk GS, 131I Two farms / region Quarterly 90SR(Q), 137 Cs Four regions Monthly (131I weekly >

GS , 1 Cs(Q) Urcomposited samples f rom four 90Sr(Q), 131I nearest dairy f arms. during the grazing season) h

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Topsoil CB, CS, From 3 fields yx 137Cs site, also 3 fields irrigated with S emi-annua lly h river water downstream of plant. p GB, GS, 1311 From 3 fielos downwind of the plant S emi-annually Vegetation site CB, GS, 137Cs From 3 fields irrigated by river Annually (at harvest)

Edible Cultivated Crops water downstream from the plant. Sample corn from the cornfield having the poorest anticipated atmospheric disper-l l sion (X/Q).

Air Samples GB GS (M) Meteorological Station at plant and 7 Weekly (filters) 131 I other stations located within 15-mile radius of plant.

Air Samples Beta Gamma Same locations as filters plus 6 on- Every 4 weeks (film badge) site stations 170 l 3.8/4.8 REV

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Table 4.8.1 '

(Continued) . .'

Air Samples Camma Same location as filters plus 6 Every 4 weeks (with one (TLD) on-site stations day down for calibration)

Small Came Camma Spectra Analysis One location withi:n a one-mile radius Semi-annually Animal of flesh and liver of the plant and a background location at a distance of between 10 and 20 miles from the site.

3TES:

90 If Sr concentrations in any of the environmental media monitored increase to twice that of the preoperational period, 90S r analyses shall be conducted on the.following samples at the indicated frequency until it has been lemonstrated that the increase is not attributable to plant operations:

Sample 90Sr Analysis Frequency River Water Quarterly Lake Water h

Quarterly , O Well Water Quarterly

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. Lake & River Bottom Sediment Semi-annually h Plankton, Algae or Insects S emi-annually >

Aquatic Vegetation Quarterly Fish Quarterly Milk Monthly Topsoil Semi-annually Agricultural Crops Annually at harves'.

20DL:G SYSTEM:

B -gross beta

S -gamma scan

[M)-monthly

'Q)-quarterly l

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.8/4.8

Bases _:

I 4.8 Environmental Monitoring Program from a certain emission from It is recognized that a precise determination of environmental doseSuch infomation will be providad by the the stack is only possible by direct measurement. If the stack emission ever environmental monitoring program conducted at and around the site.such measurements will provide a is of any reaches a level such that it is measurable in the environment, basis for adjusting the proposed stack limit long before the effect in the environmentis important > to r In this regard, it t ery large g concern for oemissible dose. rate over a period of one calendar year as permitted by 10nCFR Part ~

safety margin between conditions, at any one instant g dose of interest. 4 172l Next page is 180 ,

3.8/4.8 BASES

INTERIM TECHNICAL SPECIFICATIONS TS B.2.4-RADIOACTIVE EFFLUENTS MONTICELLO NUCLEAR GENERATING PIANT UNIT 1 License No. DPR-22 Docket No. 50-263 March 1, 1976

E TABLE OF CONTENTS Page 2.4 Radioactive Effluente TS B.2.4-1 2.4.1 Specification for Liquid Waste Effluents TS B.2.4-2 2.4.2 Specifications for Liquid Waste Sampling

& Monitoring TS B.2.4-3 Bases- Liquid Wastes TS B.2.4-4 2.4.3 Specifications for Gaseous Waste Ef fluents TS B.2.4-7 2.4.4 Specifications for Gaseous Waste Sampling & Monitoring TS B.2 .4-12 Bases- Gaseous Wastes TS B.2.4-19 2.4.5 Specifications for Solid Waste Handling & Disposal Ts B.2.4-19 Bases- Solid Wastes TS B.2.4-19 i

TSB.2.4 i

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'm LIST OF TABLES & FIGURES TABLE TS E.2.4-1 Radioactive Liq.2id Waste Sampling & Analysis TS B.2.4-2 Radioactive Gaseous Waste Sampling & Analysis TS B.2.4-3 Location o f Liquid Effluent Monitors & Samplers TS B.2.4-4 Location of Gaseous Process

& Effluent Monitors & Samplers TS B.2.4-5 Gamma & Beta Dose Factors FIGURE TS B.2.4-1 Off Gas Storage Tank Gross Activity Limits TS B . 2. 4 -i l

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. 2 . ,4 LIMITING CONDITIONS FOR OPERATION Radioactive Effluents Objective: To define the limits and conditions for the controlled relcase of radioactive materials in liquid and gaseous effluents to the environs to ensure that these releases are as low as practicable. These releases should not result in radiation exposures in unrestricted areas greater than a few percent of nacural background exposures. The concentrations of radioactive materials in effluents shall be within the limits specified in 10 CFR Part 20.

To ensure that the releases of radioactive material are as low as practicable, the following design objectives apply until Technical Specifications are issued in accordance with the recently adopted Appendix I to 10 CFR Part 50:

For liquid wastes:

a. The annual dose above background to the total body or any organ of an individual should not exceed 5 mrem in an unrestricted area.

b. The annual total quantity of radioactive materials in liquid waste, excluding tritium and dissolved gases, discharged should not exceed 5 C1.

For gaseous wastes:

c. The annual total quantity of nobic gases above background discharged f> om j the site should result in an air dose due to gam.a radiation of less than 10 mrad, and an air dose due to beta radiation of less than 20 mrad, at any location near ground level which could be occupied by individuals at or beyond the boundary of the site,

d. The annual total quantity of all radiciodines and radioactive material in particulate forms with half-lives greater than eight days above background, should not result in an annual dose to any organ of an individual in an unrestricted area from all pathways of exposure in excess of 15 enu.

c. The annual total quantity of iodine-131 discharged should not exceed 1 C1.

TS B. 2. 4- 1 l

2 ..L .1 Specifications for Liquid Waste Ef fluents

a. The concentration of radioactive caterials released in liquid waste effluents shall not exceed the value specified in 10 CFR Part 20, Appendix B, Table II, Column 2, and Notes thereto, in the condenser cooling water discharge canal.

b. The cumulative release of radioactive materials in liquid waste effluent, excluding tritium and dissolved gases, shall not exceed 10 Ci in any calendar quarter,

c. The cumulative release of radioactive materiala in liquid waste effluents, excluding tritium and dissolved gases, shall not exceed 20 Ci in any 12 consecutive months.

d. The equipment installed in the liquid radioactive waste system shall be maintained and shall be operated to process radioactive liquid wastes prior to their discharge.

e. The maximum radioactivity to be contained in any liquid radwaste tank that can be discharged directly to the environs shall not exceed 10 C1, excluding tritium and dissolved gases,

f. If the cumulative release of radioactive materials in liquid effluents, excluding tritium and dissolved gases, exceeds 2.5 Ci/ calendar quarter, the licensee shall make an investigation to identify the causes for such releases, define and initiate a program of action to reduce such releases to the design objective levels listed in Section 2.4, and report these actions to the NRC within 30 days from the end of the quarter during which the release occurred.

TS B. 2.4-2 l

g. An unplanned or uncontrolled of fsite release of radioactive materials in liquid effluents in excess of 0.5 curies requires notification.

'Ihis notification to the NRC shall be made within 30 days.

2.4.2 Specifications for Liquid Waste Sampling and Monitoring a.

Plant records shall be maintained of the radioactive concentration and volume before dilution of liquid waste intended for discharge and the average dilution flow and length of time over which each discharge occurred. Sumaries of the quantities of releases shall be included in the Semi-Annual Radioactive Ef fluent Report. Estimates of the sampling and analytical errors associated with each reported value shall be included,

b. Prior to release of each batch of liquid waste, a sample shall be taken from that batch and analyzed for the concentration of each significant gamma energy peak in accordance qith Table 2.4-1 to demonstrate compliance a with Specification 2.4.1 using che flow rate into which the waste is discharged during the period of discharge.

c.

Sampling and analysis of liquid radioactive vaste shall be perfome'd in accordance with Table 2.4-1. Prior to taking sampics from a monitoring tank, at least two tank volumes shall be recirculated,

d. The radioactivity in liquid wastes shall be continuously monitored and recorded during release. Whenever these monitors are inoperable for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, two independent samples of each tank to be discharged shall be analyzed and two plant personnel shall independently check valving prior to the discharge. If these monitors are inoperable for a period exceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, no release from a liquid waste tank :: hall be made and any release in progress shall be terminated.

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l TS B. 2. 4-3

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e. The flow rate of liquid radioactive vaste shall be concinuously measured and recorded during release,

f. The continuous mor.itors listed in Table 2.4. 3 shall be calibrated at least qunrterly by means of a solid radioactive source which has been calibrated to a National Bureau of Standards source. Each monitor shall also have a functional test monthly and an instrument check prior to making a release.

g. During each relee se of liquid radioactive vaste, the liquid radvaste discharge monitor readings shall be correlated with the results of analyses performed prior to release.

Bases: These Specifications are applic.able until Specifications prepared in accordance with Appendix I to 10 CFR Part 50 are issued by the Cocunis sion. In some cases these Specifications may be more restrictive than required by Appendix I. In the event that plant availability is adversely affected by these Specilications, the licensee may apply to the Commission for appropriate Technical Specification changes on a case by case basis.

Specification 2.4.1.a requires the licensee to limit the concentration of radioactive materials in liquid waste effluents released from the site to levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for unrestricted areas. This specification provides assurance that no member of the general public will be exposed to liquid containing radioactive materials in excess of limits considered permissible under the Commission's Regulations. 1 TSB.2.4-4

t Specifications 2.4.1.b and 2.4.1.c establish the upper limits f or the release of radioactive materials in liquid effluents. The intent of these Specifications is to pemit the licensee the f b ibility of operation to assure that the public is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the 1cvels nomally achievabic when the plant and the liquid waste treatment systems are functioning as designed. Releases of up to these levels will result in concentrations of radioactive material in liquid waste effluents at small percentages of the limits specified in 10 CFR Part 20.

Specification 2.4.1.d requires that the licensee maintain and operate the equipment installed in the liquid waste systems to reduce the release of radioactive materials in liquid effluents to as low as practicable con-sistent with the requirements of 10 CFR Part 50.36a. Normal use and maintenance of installed equipment in the liquid waste system provides reasonable assurance that the quantity released will not exceed the design obj ec tive. In order to keep releases of radioactive materials as low as l practicable, the specification requires operation of equipment whenever a release of radioactive liquid vaste is made.

Specification 2.4.1.e restricts the amount of radioactive material that could be inadvertently released to the environment to an amount that will not exceed the Technical Specification limit.

TS B. 2. 4- 5

In addition to' limiting conditions for operation listed under Specifications

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2.4.1.b and 2.4.1.c, the reporting requirements of Specification 2.4.1.f delineate that the licensee shall identify the cause whenever the cumulative release of radioactive materials in 1+. quid waste effluents exceeds one-half the design objective annual quantity during any calendar quarter and describe the proposed pt. gram of action to reduce such releases to design objective levels on a timely basis. This report must be filed within 30 days follow, he calendar quarter in which the release occurred.

Specification 2.4.1.g provides for reporting spillage or releese events which, while below the limits of 10 CFR Part 20, could result in releases higher than the design objectives.

The sampling and monitoring requirements given under Specification 2.4.2 provide assurance that radioactive materials in liquid wastes are properly controlled and monitored in conformance with the requirements of Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50 and peemit the licensee and the Commission to evaluate the plant's performance relative to radio-active liquid wastes released to the environment. Reports of the quantities of radioactive materials released in liquid waste effluents are furnished to the Commission semi-annually. On the basis of such reports and any additional information the Commission may obtain from the licensee or others, the Cannission may from time to time require the licensee to take such action as the Commission deems appropriate.

The points of release to the environment to be monitored in Section 2.4.2 include all the monitored release points listed in Table 2.4-3.

TS B. 2. 4-6

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,. . 2 . ,4. 3 Specifications for Caneous Waste Effluents The terms used in these Specifications are as follows:

subscripts y, refers to vent releases e, refers to stack releases 1, refers to individual noble gas nuclide (Refer to Tabic 2.4-5 for the noble gas nuclides considered)

QT = the total noble gas release rate (Ci/sec)

= [Qi sum of the individual noble gas radionuclides detemined to be present by isotopic analysis

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K = the average total body dose factor due to gamt. emission (rem /yr per Ci/sec)

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L = the average skin dose factor due to beta emissions (rem /yr per Ci/sec)

II = the average air dose factor due to beta emissions (rad /yr per Ci/sec)

R = the average air dose factor due to gama emissions (rad /yr per C1/sec)

The values of E, 5, E, and 5 for the vent relencen are determined from the isotopic a.alysis perfomed at the discharge.of the steam jet air ejectors ac delinented in Epecification 2.h.h.c. The vslues of E, L, E, and 5 for the stack are detemined from the isotopic analynis perfomed at a point prior to dilution and discha"Ee na delineated in Specification 2.h.4,c. hw values chould be determined each time isotopic ana2ysis is required as followu:

l 5= (1/QT) EQK ii 1

5= (1/QT) QiL i E= (1/QT) CQM ii 1

5= (1/Q)T CS Nii i

'1BB.2. b 7

where the values of K t , L 1, Mg and Ni are provided in Table 2.4-5, and are site dependent ga::na and beta dose f actors.

Q = the measured release rate (Ci/sec) of the radiciodines and radioactive materials in particulate fonns with half-lives greater than eight days.

1.1 = a factor accounting for the increased stopping power of tissue relative to air for beta radiation

a. Should any of the conditions of 2.4 3.a(1) or (2) be exceeded, the licensee chall take appropriate corrective action to bring the releases within these limits.

(1) The release rate limit of noble gases frera ti alte shall be such that 2.0 (Q Tv + II v

S Ts s }

and 0.33 (Ly + 1.1N ) + qts (L s + 1.1Ny 1, 1 (2) The release rate limit of all radioiodines and radioactive materials in particulate form with half-lives greater than eight days, released to the environs as part of the gaseous vester from the site shall be such that 5 4 3.7 x 10 Qy + 2.5 x 10 q -<, 1 s

TSB.2.4-8 1

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• . . 'b. Should any of the conditions of 2.4.3.b (1), (2), (3), (4), (5), or (6) be exceeded, the licensee shall take appropriate corrective action to bring the releases within these limits.

(1) The average release rate of noble gases fr m the site during any calendar quarter shall be such that 13(QTvEv + Q5)

Ts s

- 1 and 6.3( +

Q Tv ETs s ) 5 l (2) The average release rate of noble gases from the site during any 12 consecutive months shall be

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25(Q Tv v 9Ts s) b]

and 13( Q Tv v + qts s) S1 (3) The average release rate per site of all radioiodines and radio-active materials in particulate form with half-lives greater than eight days during any calendar quarter shall be such that 13 ( 3.7 x10 Q y + 2.5 x10 Qs ) b. 1 (4) The average release rate per site of all radiciodines and radioactive materials in particulate form with half-lives greater than eight days during any period of 12 consecutive months shall be such that 25 ( 3.7x10 Qy +

2.5x10' Qs) b1 (5) The amount of iodine-131 released during any calendar quarter shall not exceed 2 Ci.

(6) The amount of iodine-131 released during any peried of 12 consecutive months shall not exceed 4 C1.

I TS B. 2. 4 -9

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c. Should any of the conditions of 2.4.3.c.1), (2), or (3) below exist, the licensee shall make an investigation ta identify the causes of the release rates, define and initiate a program of action to reduce the release rates to design objective levels listed in Section 2.4 and report these actions to the NRC within 30 days from the end of the quarter during which the releases occurred.

(1) If the average release rate of noble gases from the site during any calendar quarter is such that 50 (Q T v + 9 TsE s ) > 1 or 25 (Q Tv v + S ) > 1 Ts s (2) If the average release rate per site of all radiciodines and radioactive materials in particulate form with half-lives greater than eight days during any calendar quarter is such that 5 0 50 ( 3.7x10 Q + 2.5x10 Q,) >l (3) If the amount of iodine-131 released during any calendar quarter is greater than 0.5 C1.

TS B. 2. 4- 10

., , d. Whenever gaseous vastes are being released frce the offgas treatment system, at least one Main Stack monitor shall be operable and set to alarm and initiate automatic termination of offgas discharge prior to exceeding the limits of Specification 2.4.3.a above. The capability of each autcuatic isolation valve shall be demonstrated quarterly. If no Main Stack monitor is operating, releases from the offgas system shall be terminated within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

e. During power operation the Reactor Buildf ng Ventilation System monitoring system shall be operable and set to alarm and initiate autcnatic termina-tion of Reactor Building ventilation air discharge prior to exceeding the limits of Specification 2.4.3.a above. The capability of automatic isolation of the ventilation system shall be demonstrated quarterly.

If the Reactor Building Ventilation System monitoring system is not operating, releases frce the Reactor Building Ventilation System shall be teminated within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

f. If the gross radioactivity release rate of noble gases at the steam jet air ejector monitors exceeds, for a period greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the equivalent of 260,000 uCi/see following a 30-minute decay, notify the NRC within ten days,

g. The drywell shall be purged through the standby gas treatment system.

h. Except as specified in Specification 2.h.31 belov, at least two hydrogen monitors in the offgas line downstream of each operating recombiner shall be operable during power opere. tion. If the hydrogen concentration reaches a set point of four percent by volume, the offges flow shall be stopped automatically by closing the valves upstream of the af fected recombiner.

1. Whenever the required hyd.rogen monitors are not operable, offcas flow to the compresced storage subsystem shall be terminated.

j. The maximum gross radioactivity contained in one gas decay tank after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> holdup that can be discharged directly to the environs shall be less than 22,000 curies of Xc-133 dose equivalent.

k. The mechanical condenser vacuum pump shall be capable of being isolated and secured on a signal of high radioactivity whenever the main steam line isolation valves are open or it shall be isolated.

1. At least once during each operating cycle automatic isolation of the mechanical condenser vacuum pump shall be verified.

m. An unplanned or uncontrolled offsite release of radioactive materials in gaseous effluents in excess of 5 curiec of noble gas or 0.02 curie of radiciodine in any one hour period shall be reported within 30 days.

2.4.4 Specifications for Gaseous Waste Sampling and Monitoring

a. Plant records shall be maintained of the sampling and analyses results. Summaries of the quantities of releases shall be included in the Effluent and Waste Disposal Semiannual Report. Estimates of the sampling and analytical error associated with each reported value should be included.

b. Prior to opening a Turbine Building roof vcnt, except in the event of fire, and daily thereaf ter, air samples shall be taken in three locations shown to be representative of radiction levels on the operating floor and analyzed for radioactive iodines and particulates with half-lives greater than eight days. In addition, when a vent is open a continuous radioactive iodine anc' particulate air monitor shall be in operation on the turbine floor. The average concentration of radioiodine and particulate matter measured in the Turbine Building shall be used in conjunction with the design flow rate of the roof exhausters in determing Turbine Building release rates. These release rates shall be included in the "Qv" terms in Specifications 2.4.3.a, 2.4.3.b, and 2.4.3.c. The position of the Turbine Building roof vents shall be logged once each shift.

TS B. 2. h-12

l l

j -

• • . c. An isotopic analysis shall be made of a representative sample of gaseous activity at the discharge of the steam jet air ejectors and at a point prior to dilution and discharge.

(1) at least monthly, and (2) following each refueling outage, and (3) if the steam det air ejector monitors indicate an increase of greater than 50% in the steady state fission gas release after factoring out increases due to power changes.

Following each analysis, steam jet air ejector monitor and stack monitor readings shall be correlated with the results of the analysis,

d. The continuous monitors listed in Table 2.4-4 shall be calibrated at least quarterly by means of a known solid radioactive source which has been calibrated to a National Bureau of Standards source. Each monitor shall have a functional test at least monthly and an instrument check at least daily.

e. Sampling and analysis of radioactive material in gaseous ' .ste, including

- particulate forms and radiciodines shall be performed in accordance with Table 2.4-2.

f. The hydrogen monitors shall be functionally tested monthly and calibrated quarterly with an appropriate gas mixture source. Each monitor shall have a sensor check at least daily.

g. Condenser air inicakage shall be evaluated weekly and used in conjunction with steam jet air ejector offgas isotopic analyses and Figure 2.4-1 to to determine that the limit of Specification 2.4.3.j is not exceeded.

TSB.2.4-13

l l

i . .

Bases: The release of radioactive materials in gaseous waste ef fluents to unrestricted areas shall not exceed the concentration limits specified in 10 CFR Part 20 and should be as low as practical in accordance with the re-quirements of 10 CFR Part 50.36a. These specificat1ons provide reasonable assurance that the resulting annual air dose from the site due to ga:mna radiation will not exceed 10 mrad, the annual air dose from the site due to beta radiation will not exceed 20 mrad from noble gases, that no individual in an unrestricted area will receive an annual dose to the total body greater than 5 mrem or an annual skin dose to the total body greater than 15 mrem from fission product noble gases, and that the annual dose to any organ of an individus1 from radiciodines and radioactive material in particulate form with half-lives greater than eight days will not exceed 15 mrem from the site.

At the same time these specifications permit the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided with a dependable source of power under unusual operating con-ditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20. Even with this operational flexibility, the annual releases will not exceed a small fraction of the concentration limits specified in 10 CFR Part 20.

The design objectives have been developed based on operating experience taking into account a combination of system variables including defective fuel, primary system leak, age, and the performance of the various waste treatment systems.

l TS B. 2. 4- 14

Specification 2.4.3.a(1) limits the release rate of noble gases from the site so that the corresponding annual ga::na and beta dose rate above back-ground to an individual in an unrestricted area vill not exceed 500 mrem to the total body or 3000 mrem to the skin in ecupliance with the limits of 10 GFR Part 20.

For Specification 2.4.3.a(1), ga::na and beta dose factors for the individual noble gac radionuclides have been calculated for the plant gaseous release points and are provided in Table 2.4-5. The expressions used to calculate these dose f actors are based c:. dose models deriwd in Section 7 of Meteorology and Atomic Energy-1968 and model techniques provided in Draf t Regulatory Guide 1.AA.

Dose calculations have been made to determine the site boundary location with the highest anticipated dose rate from noble gases using on-site meteorological data and the dose expressions provided in Draft Regulatory Guide 1. AA. The dose expression considers the release point location, building wake effects, and the physical characteristics of the radionuclides.

The offsite location with the highest anticipated annual dose from released noble gases is 700 meters in the SSE direction.

The release rate Specifications for radiciodine and radioactive material in particulate form with half-lives greater than eight days are dependent on existing radionuclide pathways to man. The pathways which were examined for these Specifications are: 1) individual inhalation of aftborne radionuclides,

2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, and 3) deposition onto grassy areas where milch animals graze with consumption of the milk by man, which was determined to be the most limiting pathway. Methods for estimating doses to the thyroid via these pathways are described in Draf t Regulatory Guide 1.AA.

TS B. 2. 4- 15

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

Th2 ot'isite location with the highest enticipated thyroid does rats from radioiodines and radioactive material in particulate form with half-lives greater than eight days was determined using on-site meteorological data and the expressions described in Draft Regulatory Guide 1.AA.

Specification 2.4.3.a(2) limits the release rate of radiciodines and radioactive taterial in particulate form with half-lives greater than eight days so test the corresponding annual thyroid dose via the most restrictive pa:hway is less than 1500 mrem.

For radiciodines and radioactive material in particulate ivna with half-lives greater than eight 6:n , the most restrictive location is a dairy farm located 3700 metars in the NNE direction (vent X/R = 4.3x10'7sec/m3 ; stack

~

X/Q = 2.5x10 sec/m ).

Specification T. 4.3.b establishes upper offsite levels for the releases of nobic gases and radioiodines and radioactive material in particulate form with half-lives greater than eight days at twice the design objective annual quantity during any calendar quarter, or four times the design objective annual quantity during any period of 12 consecutive months. In addition to the limiting conditions for operation of Specifications 2.4.3.a and 2.4.3.b, the reporting requirements of 2.4.3,c provide that the cause shall be iden!ified whenever the release of gaseoua effluents exceeds one-half the design objective annual quantity during any calend1r quarter and that the proposed program of action to reduce such release t< tes tc the design objectives shall be described.

Specification 2.4.3.d and 2.4.3.e are in accordance with Design criterion 64 of Appendix A to 10 CFR Part 50.

I TS B. 2. 4- 16

., . - Specification 2.4.3.f is intendsd to monitor tha performance of the core. An

, , inerees- a the activity levels of gaseous releases may be the resu' el defective fuel. Since core performance is of utmost importance in the result-ind doses from accidents, a report must be filed within 10 Jays following the specified increase in activity level at the steam jet cir ejector.

Specification 2.4.3.g requires ti.at the drywell atmosphere receive treatment for the removal of gaseous iodine and particulates during purging.

Specification 2.4.3.h requires that hydrogen concentration upstream of the compressed radioactive gaseous storage tanks shall be monitored whenever the comprected storage subeystem is in uce.

Specirleation 2.4 31 requires offgas flow to the compressed storage tankr to be terminated in the event that the hydrogen monitors downstream of the recombiners are inopernble. This prevents the poccible accumalation of an explosive mixture in portions of the offgas system which are not designed to fully withstand a hydrogen detonction.

Specification 2.4.3.j limits the maximum gross activity in one decay tank on the basis that accidental release of its contents to the environs by operator error after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> decay should not result in exceeding the dose equivalent to the maximum quarterly release rate specified in Specification 2.4.3.c.1. Staff analysis of an elevated release under accident meteorology for a minimaa release period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> indicated a release of 22,000 curies of Xe-133 or the dose equivalent would result in an air dose from the site of 20 mrad from noble gases.

Calculations have been perfcrmed to determine the relationship between steam jet air ejector offgas activity and composition and condenser air inicakage.

These calculations were used to determine the curves presented in Figure 2.4-1.

The results of the measurement of condenser air inleekage and the average air ejector offgas release rate are used in conjunction with the TS B. 2. 4- 17 l

. most recent offgas isotopic analysis to determine if the maximam permitted Xe-133 dose equivalent tank radioactivity contents may be exceeded. This analysis is adequate to initiate corr,ective action in the unlikely event that che tank radioactivity limit is being approached.

fpecifications 2.4.3.K and 2.4.3.1 require that the mechanical vacuum pump be provided with automatic isolation capability to limit the release of activity from the main condenser during an accident.

Specification 2.4.3.m provides for reporting release events which, while below the ILmits of 10 CFR Part 20, could result in releases higher than the design objectives.

The sampling and monitoring requireme .ts given under Specification 2.4.4 provide assurance that radioactive material released in gaseous waste effluents are properly controlled and monitotes in conformance with the requirements of Dasign Crite'ria 60 and 64. These requirements provide the data for the licensen and the Commission to evaluate the plant's performance relative to radioactive vaste effluents released to the environment.

Reports on the quantities of radioactive materials released in gaseous effluents are furnished to the Commission semiannually. On the basis of such reports anj any additional information the Commission may obtain from the licensee or others, the Commission may from time to tbne require the licensee to take such action as the Commission deems appropriate.

The points of release to the environment to be monitored in Section 2.4.4 include all the wonitored release points as provided for in Table 2.4-4.

These Specifications are applicable for the interim period until the date that Specifications prepared in accordance with new Appendix I become effective. In some cases these Specifications may be more restrictive than TS B. 2. 4-18

I v

required by Appendix 1. In the event that plant availability is adversely affected by these specifications, the licensee may apply to the Counission for appropriate Technical Specification changes on a case by case basis.

2.4.5 Specifications for Solid Waste Handling and Disposal

a. Measurements shall be made to determine or estimate the total curie quantity and principle radionuclide composition of all radioactive solid waste shipped offsite.

b. Sunnaries of radioactive solid waste shipments, volumes, principal radionuclides, and total curie quantity, shall be included in the Effluent and Waste Disposal Semiannual Report.

Bases: The requirements for solid radioactive vaste handling and disposal given under Specification 2.4.5 prov'ie assurance that solid radioactive materials stored at the plant'and shipped offsite are packaged in confonaance with 10 CFR Part 20,10 CFR Part 71, and 49 CFR Parts 170-178.

I TS B. 2. 4- 19

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

t

, , TABLE TS B.2.4-1 RAb10 ACTIVE L1 QUID WASTE SAMPLISG AND ANALYSIS Liquid Detectable Sampling Type of Concentrations Source Frequency Activity Analysis (uCi/ml)"

A. Monitor Tank Releases Each Batch Principal Gama Emitters 5 x 10*N One Batch / Month Dissolved Cases

• 10'E Weekly Compositec Ba- La- 140, 1- 131 10-6 Monthly CompositeC Gr-89 5 x 10-8 11 - 3 10-5 G4 osc A%ha 10-7 Quarterly ComposM Er-90 5 x 10-8 B. Primary Coolant d Weekly 1-131, 1-133 10-6 a

The detectability limits for activity analysis are based on the technical feasibility and on the potential significance in the environment of the quantities releaseJ. For some nuclidea, lower detection limits may be readily achievable, and when nuclides are b measured below the stated limits, they should also be reported, For certain mixtures ef gana emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the same in much greater concentrations. Under these circumstances, it will be more appropriate to calculate the concentrations of such radionuclides using measured ratios with thnse radionuclides which are routinely identified and measured, cA compositt sample is one in which the quantity of liquid sampled is proportional to d

the quantity of liquid waste discharged.

The power level ard cleanup or purification flow rate at the cample time shall also be reported.

'For disco 1ved noble gases in water, assume a stPC of 4 x 10-5uci/ml of water.

TABLE TS B.2.4-1

a+

TABLE TS B.2.4-2 RADICACT"/E GASSOUS WASTE SAMPLITU ARD ANALYSIS Gaseous Sampling Type of Detectable Source Frequency Activity Analysis Concentratior.s (uCi/ml)*

A. Containment Purges Ihch Purge Principal Gam Emitters 10-4 b 11 - 3 10-6 B. Environmr ntal Felease Monthly Principal Gterna Fritters I 10-4 b,c Points (CasSnmpleb) 11- 3 10-6 Weekly I.131 10-12 d

( Charcoal Eamples )

Monthly I.133, I.135 10-10

( Charcoal Samples )

Weekly Principal Gamma Ihitters 10-11 d (Particulates) (at least for Ba-la-140 and I-131)

Monthly Composite

• Br-89 10-11

( Particulates )

l Gross Alpha 10-11 Qtrly Composite *

(Particulates) E r-W 10 4

" 'Ibe above detectability limits for activity analysis are based on technical feasibility anc on the potential significance in the environment of the quantitiec released. For some nuclides, lower detection limits may be readily achievable, and when nuclides are measured below the stated limits, they should also be reported.

b Analyses shall also be performed following each refueling, startup, or similar operational occurrence which could alter the mixture of radionuclides.

c For certain mixtures of gamma emitters, it may not be possible to measure radionuclides at levels near their sensitivity limits when other nuclides are present in the sample at much higher levels. Under these circumstances, it will be more appropriate to calculate the levels of such radionuclides using observed ratios with those radionuclides which are measurable.

l l TABLE TS B.2.4-2 (Page 1 of 2)

v ,

IABLE TS B.2.4-2 Notes (continued) d When the average daily gross radioactivity release rate exceeds that give in 2.4.3.c(1) or where the steady-state gross radioactivity release rate increases by 50% over the previous corresponding power level steady-stat release rate, the iodine and particulate collection devicesfor the release -

point whose contribution exceeds 50% of these rates shall be removed and analyzed to determine the change in iodine-131 and particulate release rate.

The analyses for this release point shall be done dally following such change until it is shown that a pattern exists which can be used to predict the release rate after vbich it may revert to weekly sampling.

• To be representative of the aveutsc quantities and concentr ti radioactive materials in pat t. ct a ons of samples effluent streams.should be collectec in P.wortion to the rate of flow of theiace f

Isotopic analysis p6rformed in accordance with Specification 2.4.4.c at the discharge of the steam jet air ejectors and at a point prior to dilution and discharge of gaseous waste from the offgas system.

Concentrations of individual gamma emitters in the Reactor Building vent are expected to be below the minimum detectable levels with the exist-ing analytical equipment. Therefore, isotopic analyses of samples from the vent will not normally be performed and the isotopic content will be assumed to be that existing at the steam jet air ejector.

P TABLE TS B.2.4-2 (Page 2 of 2)

e t

.~

~

TABLE TS B 2.4-3 LOCATION OF LIQUID EFFLUENT MONITORS AND SAMPLERS High Pmeess Stream Auto Control Grab Measurement Liquid  ;

or Radiation to Isolation Continuous Sample Gross Dissolved Isotopic level l Felease Point Alarn Valve Monitor Statien Activity I Cases Alpha N-3 Analysis Alarm l Floor Drain X X X X X X X j w*

Sample Tank Laundry Drain X I X X X X X Tank Primary Coolant X X l

System ~

Liquid Fadvaste X X X Discharge Pipe i

Service *4ater X X X Dischstge Pipe Closed Cooling X X X ~'s

' dater Erstem l Discharge Canal X X X Sa=pler TABLE T' B. 2.4-3

y 0

.~. .

TABLE TS B.2.4-4 LOCATION OF CASEOUS PROCESS AND EFFLUENT MONITORS AND SAMPLERS Grab Radiation Auto Control to Continuous Sample Measurenent

, Proccos Stream or Release Point Alarm Isolation Valve Monitor Station Noble Gas I Particulate H-3 Alpha Cond:nser/ Air Ejector (before X X X X X

• gc3 treatment system)

Dffg.ra Treatment System (before X X dilution and discharge)

Main Stack X X X X X X X X X fcactor Building X X X X X X X X X jVantilation System

! I brbineBuilding X X X g boercting rioor Mechanical Vacuum

• X Pu=p

• Isolation on main steam line high radiation.

1 TABLE T5 B.2.4-4

o 4 TABLE TS B.2.4-5 -

?

CA}fR AND EETA DOSE FACTORS [

Monticello Doese Factors for Vent Dose Factors For Stack Noble Gas 'iv iv iv iv is is is is Radionuclide Total Body Skin Beta Air Ca=ma Air Total Body Skin Beta Air Ganma Air rem /yr rem /g rad /vr rad /vr rem /vr rem /vr red /vr red /yr Ci/sec CL/sec C1/see Ci/sec Ci/sec Ci/see C1/see Ci/sec fir-83m 2.0 x 10 0 1.6 0.13 2.0 x 10' O 0.063 5.3 x 10' Kr-85m 2.0 8.0 11 2.1 0.59 0.32 0.43 0.6 Kr-85 0.023 7.4 11 0.024 8.6 x 10-3 0.29 0.43 9.1 x 10' Kr-87 6.6 54 57 6.9 2.5 2.1 2.3 2.7 Kr-88 15 13 16 16 6.3 0.52 0.64 6.6 Kr-89 9.4 56 58 9.9 2.1 2.2 2.3 2.2 Xe-131m 0.69 2.6 6.1 0.89 0.15 0.10 0.24 0.18 i

Xe-133m 0.54 5.5 8.1 0.75 0.12 0.22 0.33 0.14 Xe-133 0.63 1.7 5.8 0.79 0.13 0.067 0.23 0.14 Xe-135m 3.8 3.9 4.1 4.1 1.2 j 0.16 0.16 1.3 Xe-135 2.9 10 14 3.0 0.94 0.41 0.54 0.99 Xe-137 1.1 67 70 1.2 I

t 0.25 2.7 2.8 0.26 Xe-138 93 23 26 9.8 31 0.91 1.0 32 <

TABLE TS B.2.4-5

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NRC DISTRIBUTION ron PART 50 DOCKET MATERIAL .

1 i '10M:  : ct thern Mat es Por O DATt or o cvMENT lTO Mr ft'ello t h r.nc ,y,ol i n , Mn 1-1-76 L U Mr'er*

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PLANT NAME:  ?'ent ice 11 s SAFETY FOR ACTION /INFORMATION ENVIRO J 7 t> ch!

ASSIGNED AD :

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PROJECf l.AhAGERt PROJECT MANAGER :

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INTERNAL DISTRIBUTION Eb.lllE) ._. SYSTEMS _.S AFElY PLANT _. SYSTEMS JNVInn *rrCl!

/ NRC TDR ICINEMAN TEDESCO-- __ .lKdS1

/ I &_E_f d .___SCHROEDER BENAROYA BALLARD SPANGLER

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/ GOSSICK & STAFF ENGU{EEBING_ IPPOLITO MIF.C MACCARY SITE TECll KNIG11T OPERATING REACTORS _GAF211Li, CASE SIINEIL a STEPP

_ __HANA1]ER STF4LO _ .

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' PROJECT MANAGEMENT REACTOR SAFETY / EISENiitTT SITE ANALYSIS BCYD _ _

ROSS / SilAO VOLIJiER P. COLLINS NOVAK / BAER BUNCil 110DSTON ROSZTOCtY ./ SCINEliCER J. COLLINS PETERSON CllECK / GRIMES KREGER _ _ _

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llELTEMES AT & 1 SITE SAFETY &_.E.b1TRC SKOVliOLT SALTZMAN ANALYSIS

, RlrIRERG DENTON & MULLER

, NUMBER EXTERNAL DISTRIBJTION CONT _.

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