ML18139A341

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Forwards Comments on Proposed Amend to Tech Specs.Util Must Forward Response to Comments on Offsite Dose Calculation Manual & Process Control Program Re Solidification of Radwastes within 45 Days
ML18139A341
Person / Time
Site: Surry  Dominion icon.png
Issue date: 06/04/1980
From: Varga S
Office of Nuclear Reactor Regulation
To: Ferguson J
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
References
NUDOCS 8006200521
Download: ML18139A341 (46)
v • d • e


Text

{{#Wiki_filter:--r.-_-r~~., !~ ~...-;.__-:: - Docket Nos. 50-280 and 50-281 Mr. J. H. Ferguson Executive Vice Pcesident - Power Virginia Electric and Power Company

  • Post Office Box 26666 Richmond, Virginia 23261 r

Dear M~. Ferguson:

- ~ DISTRIBUTION Docket Files 50- BOV281 NRC PDRs (2) Local PDR TERA NSIC '. NRR Reading ORBl Reading D. Eisenhut T. Novak S. Varga D. Neighbors C. Parrish I&E (3) Attorney,- OELD By letter dated March 15, 1979, you proposed to amend the:existing Technical Specifications-of Surry Power Station, Units l and_,2, for the. radiolog.ical effluent and environmental monitoring systems;':/_.'..:

  • ,Our revJew of the proposed Technical Specific~tions was bas:¢d on the model :Radiological Effluent Technical Specifications for Pressurized.

Hater Reactors. NUREG-0472, Revision 2; July.1979. Our coJnments* and a .,. mar!u:~d-up copy of your propo'sed Technical. Specifications ir:e* ~nclosed}l.S Enclosures 1 and 2, respectively. You sho*uld incorporate1fthese comments into.your proposed lechnical Specifications.

  • Y~u have not submitted an Offsite Dose Calculation Manual;. (ODCM) nor ii Process Control Program {PCP} for solidification of radioactive wastes for Surry Power Station, Units l and 2. Whether you use* a contractor for waste solidification~dewatering or perform your own was~:~ processing, a PCP should be submitted.

He request that the OOCM and* PCP be submitted for our review, and that a response to the enclosed comment~ be made within forty-five (45) days of receipt of this letter.**

Enclosures:

As Stated cc: w/o enclosures See next ~age 8 00 62 00. 521 Sincerely, il'ia1na1 *sign . 6. A. VargGted by:

??f.,

Steven A. Varga, Chief* Operat_ing Reactors Branch #1 Division of Licensing _.//J / ~/ ' L .... u.................... :........ * -t.u.s. GOVERNMENT PRINTING OFFICE: 1979-289-369

1.

Enci osure 1 e ETSB COMMENTS ON SURRY POWER STATION, UNIT NOS. 1 AND 2 RETS We have reviewed the subject radiological effluent Technical Specifications as submitted, and have marked them up to reflect a document which, subject to resolution of these comments, is acceptable to us. We have in a number of cases, changed the licensee's wording, content, and table format to make them conform to the contents of NUREG-0472, Revision 2. Specific changes made may require subsequent discussion.

2.

In Section 1.0, DEFINITIONS, add definitions for DOSE EQUIVALENT I-131, SOLIDIFI-CATION, PROCESS CONTROL PROGRAM, GASEOUS RADWASTE TREATMENT SYSTEM, VENTILATION EXHAUST TREATMENT SYSTEM, PURGE-PURGING, and VENTING; and modify definitions for OFFSITE DOSE CALCULATION MANUAL and CHANNEL FUNCTIONAL TEST as shown in markup.

3.

Modify the following specifications as shown in markup: 3.7.F, 3.11.A.1.a, 3.11.A.1.b, 3.11.A.1.c, 3.11.A.2.a, 3.11.A.2.b, 3.11.A.3.a, 3.11.A.3.b, 3.11.B.1.a, 3.11.B.l.b, 3.11.B.1.c, 3.11.B.2.a, 3.11.B.2.b, 3.11.B.3.a, 3.11.B.3.b, 3.11.B.4.a, 3.11.B.4.b, 3.11.B.5.b, 4.9.A.1.c, 4.9.A.l.d, 4.9.A.1.e, 4.9.A.3.a, 4.9.A.3.b, 4.9.B.l.c, 4.9.B.l.d, 4.9.B.2, 4.9.B.3, 4.9.B.4.a, 4.9.B.4.b, 4.9.B.5, 4.9.D, and 4.9.c.

4.

In Table 3.7-5, perform the following: a) Indicate capability for monitoring or sampling the Turbine Building (Floor Drains) Sumps Effluent Line, and indicate. provisions for termination of releases via this pathway in accordance with NUREG-0472. See also comment No. 14.

e e - b) Indicate capability for measuring flow rate in your Liquid Radwaste Effluent Line, Steam Generator Blowdown Effluent Lines, and Discharge Canal in accordance with NUREG-0472. c) Modify the Table, including the ACTION statements, as shown in the markup. d) Indicate capability for automatic termination of releases, on high activ-ity, in your Steam Generator Blowdown Effluent Lines. e) Unless an alarm/trip setpoint is based.on the Radioactivity Recorder listed as Item lrb; delete it from the Table, along with its corresponding ACTION 3 Table Notation. f) Indicate capability for indicating tank levels, for.any tanks located out-side plant buildings which contain potentially radioactiye liquids. g) Indicate capability for monitoring for explosive gas potential in the Process Vent Subsystem (i.e., H2;o2 analyzers), in accordance with NUREG-0472. Also see comment No. 7. h) Indicate capability for monitoring Steam Generator Blowdown Vent System effluent-releases in accordance with NUREG-Oa.72 *.. i) Indicate capability for monitoring the Process Vent System effluent flow rate in accordance with NUREG-0472.

5.

Provide a specification for curie content in outdoor liquid holdup tanks contain-ing potentially radioactive liquid, in accordance with NUREG-0472.

' r

6.

In Table 4.9-3: a) Indicate your sampling and analysis program for the Steam Generator Blow-down Flash Tank Vent System in accordance with NUREG-0472, or justify its absence. b) Indicate your sampling and analysis program for the Containment Purge, in accordance with NUREG-0472. c) Modify the Table, including Notation, as shown in the markup.

7.

Provide a specificati_on for an explosive gas mixture in the waste gas holdup system, in accordance with NUREG-0472. Also see comment No. 4.g.

8.

Provide a specification for Solid Radwaste System operability, in accordance with NUREG-0472.

9.

Provide a specification for Total Dose relating to compliance with 40 CFR 190, in accordance with NUREG-0472.

10.

Change the Bases for Specification 3.11 as shown in markup, and include Bases for new specifications which need to be provided, in accordance with NUREG-0472.

11.

Figure 3.11.1 is insufficiently detailed. Provide new figure(s) in accordance with instructions for Figures -;l"'-"3 and-5-;l of NUREG-0472.

T e -

12.

Delete Specifications 4.9.A.l.a, 4.9.A.l.b, 4.9.B.l.a, and 4.9.B.l.b, as shown in markup.

13. Modify Specification 4.9 Bases as shown in markup.
14. Modify Table 4.9-2 and its accompanying Table Notation as shown in markup.

In conjunction with Comment 4.a, note the sampling requirements for Turbine Building Drains if continuous monitoring is not provided for this release pathway.

15.

Modify the Basis for Radioactive Effluent Monitoring Instrume!1tation as shown in markup.

16.

In Specification 4.9.D, Table 4.1-1 is referenced for effluent monitoring instru-mentation test operations and frequencies. This table was not submitted for review, and your current Table 4.1-1 doesn 1t list any instrumentation of this type. Correct this discrepancy and ensure that information provided in this Table is in accordance with NUREG-0472.

17.

Modify Specification 6.6.3.6 as shown in markup. This is a semiannual report, not annual. NOTE: Other than Specification 6.6.3.b, no other proposed Administrative Controls sections were submitted for review. The following comments are based on your current Administrative Controls specifications.

I e 18. In Specification 6.1.C.1.e, add the following responsibilities:

11.

Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recom~endations and disposition of the corrective action to prevent recurrence to the (Superintendent of Power Plants) and to the System Nuclear Safety and Operating Committee.

12.

Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems.

19.

In Specification 6.1.C.2.i, add the following audit items:

11.

The radiological environmental monitoring program and the results thereof at least once pe_r 12 months.

12.

The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

13.

The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months.

14.

The performance of activities required by the Quality*Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977, at least once per 12 months.

20.

In Specification 6.4.A, add the following procedure requirements:

9.

PROCESS CONTROL PROGRAM implementation.

10.

OFFSITE DOSE CALCULATION MANUAL implementation. 11 Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 4.15, December 1977.

21.

Specification 6.6.1.c should be modified to read as follows: MONTHLY REACTOR OPERATING REPORT Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a

I copy to the Regiona1 Office of Inspection and Enforcement, no 1ater than the 15th of each month fo1lowing the ca1endar month covered by the report. Any changes to the OFFSITE DOSE CALCULATION MANUAL shal1 be submitted with the Monthly Operating Report within 90 days in which the change(s) was made effective. In addition, a report of any major changes to the radioactive waste treatment systems sha11 be submitted with the Month1y Operating Report for the period in which the evaluation was reviewed and accepted by the Station Nuclear Safety and Operation Committee.

22.

In Specification 6.6.2.a, add the following events as prompt notification items: (10) Offsite releases of radioactive materials in liquid and gaseous effluents which exceed the_ limits of Specification 3.11.A.l.a or 3.11.B.1.a. ( 11) Exceeding the limits in Specification* or radioactive materials in the listed tanks. shall include a schedule and a description taken to reduce the contents to within the 3.11.B.5 for the storage of The written follow-up report of activities_p1anned and/or specified limits.

23.

In Specification 6.6.2.o, add the following thirty (30) day written report items: (5) An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. The report of ah unplanned offsite release of radioactive material shall include the following information:

1. A description of the event and equipment involved.
2.

Cause(s) for the unplanned release.

3. Actions taken to prevent recurrence.
4.

Consequences of the unplanned release. (6) Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table** when averaged over any calendar quarter sampling period.

  • See Comment No. 5.
    • Reporting Levels for Radioactivity Concentrations in Environmental Samples.

1 -

24.

In Specification 6.5.B, add the following record keeping requirement:

8.

Records of analyses required by the radiological environmental monitoring program.

25.

Administrative Controls specificatjons should be added for the following, in accordance with NUREG-0472:

a.

PROCESS CONTROL PROGRAM (PCP). b, OFFSITE DOSE CALCULATION MANUAL (ODCM).

c.

MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS.

1 e other independent channels measuring the same variable.

2.

Channel Functional Test

3.

Injection of a simulated signal into an analog channel or makeup of the logic corabinations in a logic channel to verify that it is operable, including alarm and/or trip initiating action. S:~~ ~&A-ied:~o~ .i~1..to o.~ c..ki<.v---e.Ls ":;>\\....ot,,..\\.J.__ loe... tt<i; e.lose: +o +ka... s:e11-so.....- cu 'f,rc...cl, ~le.. J a.....,J._ Li.'- ~is'ta.lole... ~[ls si5 N::Ll :~ ec..tioi,,,.... sl.u,u.ld._ lo~ i v...to +k 'be1ASor-. Channel Calibration Adjustment of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the chan-nel measures. Calibration shall encompass the entire channel, including equipment action, alarm, *or trip, and shall be deemed to include ~he channel functional test.

4.

Source Check A Source Check shall be the qualitative assessment of radiation monitor response when the channel sensor is exposed to a radioactive source. H. Containment Integrity Containment integrity is defined to exist when:

1.

All non-automatic containment isolation valves, except those required for intermittent operation in the performance of normal operational activities, are locked closed and under administrative control. Non-automatic containment isolation valves may be opened intermittently for operational activities provided that they are under administrative control and are capable oi being closed immediately if required.

1 ,I ..!.. *,._;- _.I e

2.

Blind flcnges are installed where required.

3.

The equipment access hatch is properly closed and sealed.

4.

At least one door in the personnel air lock is properly closed and sealed.

5.

All automatic containment isolation valves are operable or are locked closed under administrative control.

6.

The uncontrolled containment leakage satisfied Specification 4.4. I. Reportable Occurrence

1.

Definition: Refer to Technical Specification 6.6, Station Reporting Requirements for the definitions and examples of the two categories of Reportable Occurrence Reports

a.

Prompt Notification With Written Follow-up.

b.

Thirty Day Written Reports J. Quadrant Power Tilt The quadrant power tilt is defined as the ratio of the maximum upper excore detector current to the average of the upper excore detector currents or the ratio of the maximum lower excore detector current to the average of the lower excore detector currents whichever is greater. If one excore detector is out of service, the three inservice units are used in computing the average.

1 l K. L. e ~~. 0-6 Low Power Physics Tests Low Power Physics Tests conducted below 5% of rated power which measure fundamentals characteristics of the core and related instrumentation. Fire Suppression Water System A Fire Suppression Water System shall consist of: a water source(s); gravity tank(s) or pump(s); and distribution piping with'associated sectionalizing control or isolation valves. Such valves shall in-clude yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser. M. Offsite Dose Calculation Manual (ODCM) An Offsite Dose Calculation Manual shall be a manual containing the methodology and parameters to be used in the calculatio11 of offsite doses due to.radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm/trip setpoints.G-Gnsis-eent--with the-a~~:i:eable LCG~~t~ ~-e-1'-eehn+/-ea+/--Spec i fi ca-t..:iott&.

o.

P. Q. e The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID--14844, "Calculation of Distance Factors for Power and Test Reactor Sites. 11 So [, d.;J; c.o.:t~ c"'- SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing). The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from - liquid systems is assured. A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the. purpose of reducing the total radioactivity prior to release to the environment.

  • ve.v-.til.Cl..~io~ E.xJ"'-AVst Tvru...t~ 5::!<lu.-.

A VENTILATION EXHAUST TREATMENT SYSTEM is any sys tern designed and installed. to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

s.

Pv.vt3e. - Pw--s i ":5 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. 1. v*c:A-~t i :':::5 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. ~: ~

  • .1
    • 1

I e e

3. 7 INSTRUl*lt~ITAIIJ:1 SYSTEMS Operational Safetv Instru.;;entation Applicability:

Applies to reactor and safety features instrumentation systems. Objectives: To provide for automatic initiation of the Engineered Safety Features in the event that principal process variable limits are exceeded, and to delineate the conditions of the plant instrumentation and safety cJrcuits necessary to ensure reactor safety. Specification A. For on-line testing or in the event of a subsystem instrumentation channel failure, plant operation at rated power shall be permitted to continue in accordance with TS Tables 3.7-1 through 3.7-3. B. In the event the number of channels of a particular subsystem in service falls below the limits given in the column entitled Minimum Operable Channels, or Minimum Degree of Redundancy cannot be achieved, operation shall be limited according to the requirement shown in Column 4 of TS Tables 3.7-1 through 3.7-3. C. In the event of subsystem instrumentation channel failure permitted by specification 3.7-B, TS Tables 3.7-1 through 3.7-3 need not be observed during the short period of time the operable subsystem channels are tested where the failed channel must be blocked to prevent unnecessary reactor trip.

D. The Engineered Safety Features initiation instru~entation setting limits shall be as stated in TS T~ble 3.7-4. E. Automatic functions operated from radiation monitor alarms shall be as stated in TS Table 3.7-5. F. The radioactive effluent process and monitoring instrumentation Basis OPE fLPr f:,L£ o+/- o...Ll -t-;......_e_s channels shown in Table 3.7-5 shall be ~~Awith their alarm/trip setpoints set to ensure that the limits of Specifications 3.11.A.l and 3.11.B.1 are not exceeded. ~ ()J.a.n.*,.. /-fr:() sd:flc:;,.z'Cs: ot ~<?.- ~ s!Ac.Ll be.. cldlU'" ~ (...,._ ~~ w: +t,.... +k... 0 DC...l'A,

1.

With a radioactive effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of Specifications 3.11.A.l and B.l are met, immediately suspend the release of radioactive effluents monitored by the affective channel or declare the channel inoperable.

2.

With one or more radioactive effluent monitoring instrumentation channels inoperable, take the Action shown in Table 3.7-5. -ma-i-n-t-a4:n~ Instrument Operating Conditions During plant operations, the cosplete instrumentation systems will normally be in service. Reactor Safety is provided by the Reactor

~ *. -J the differential pressure expe_cted in the event of-a large stea;n line break accident as shown in the safety analysis.3

5.

The high steam line flow differential pressure setpoint is con-stant at 40% full flow between no load and 20% load and increasing* linearly to 110% of full flow at full load in order to protect against large steam line break accidents. The coincident low T avg setting limit for SIS and steam line isolation initiation is set below its hot shutdown value. The coincident steam line pressure setting limit is set below the full load operating pressure. The safety analysis shows that these settings provide protection in the event of a large st~am line break.3 Automatic Functions Operated from Radiation-Monitors The Process Radiation Monitoring System continuously monitors selected lines containing or possibly containing, radioactive effluent. Certain channels in this system actuate control valves on a high-activity alarm signal. Additional information on the Process Radiation Monitor-ing System is available in the FSAR.4 Radioactive Effluent Monitoring Instrumentation The Radioactive Effluent Monitoring Instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid and gaseous effluents, during actual or potential

t 3. 7-8 releases. The alarui/trip setpoints for these instruments are calculated in accordance with the methodology contained in the ODCM, to ensure that the alarm/trip will occur prior to exceeding OPc./Ui 61Ltrf the limits of 10 CFR 20. The -r.!i:H.~iHineeAand use of this instru-mentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50. "T"~s ~~$+v~~t *:o"' also 1*~,c.Lv-.d..u pvov,s:iov,..s f.,v M..clh.i.+on~ {_a,,.)._ C.o.,.:lvoUJ~) +i-.~ ern-..u__.,,_~ o-t po+e+--+/-:o.L~ Uf [o.s/vt.- 5o.s M.t"xt1,.1.ve2J 1*...,, ii,..e.. (0c..sk <jC...: ~lctl>-f s*pf4A<,,..), Reference (1) FSAR - Section,7.5 (2) FSAR - Section 14.5 (3) FSAR - Section 14. 3. 2 (4). FSAR - Section 11. 3. 3

TABLE 3.7-5 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT MINIMUM CHANNELS OPERABLE APPLICABILITY ACTION l; Liquid Radwaste Effluent Line

a.

Gross Radioactivity Monitor (a) 1 ,-r 1

  • g:_. co1.""'G.J-r 4-, e, -
2.
3.
4.
s.
6.
7.
8.

~dioactivity Recorder

~"'"

Component Coolin~ Wat~ Mo_n_i_t-or--(b_) ___ l-----------:-&ee-IP::--H*:~J:1=&~r.4-~ * -NA:?- Steam Generator Illowclown Gross Activity Monitor Process Vent System

a.

Noble Gas Activity Monitor (c)

b.

Particulate & Io<l:lne Sampler (c). Containment Purge System 1 1 1

a.

Noble Gas Activity Monitor (d) 1

h.

Particulate Monitor (d) 1 -e.--Man-=i-1~&tor-&r*un-e-A-1~ea-rle11-i-t-or-s-fe-)--:l-Ven t ila t ion Vent System

a.

Noble Gas Activity Monitor h, Particulate & Iodine Sampler

c.

Effluent System Flow Monitor

d.

Sarr1pler Flow Rate Monitor Condenser Air Ejector Gross Activity Monitor (f) Component Cooling Heat'Exchanger Service Writer Monitor 1 1 1 1 1 1

  1. -See-!fs.-3-.-l-0-&-4.9-
  • -See-I'-s-3.-l0-&-4-...Se~S-Jrio.-&-4. *
  • B1:n1-:i:ng--1:-el-eases--v+/-a-1:-M-s--pa-thway-(-See-t-echn-+/-etrl-Spec--l-f+/-c-at:io~-4.-9-}-

1;;8 6 -Ntt? -NA-l, -NA-5 6 4 4 5 a *. This monitor (RM-LW-108) automatically closes effluent discharge valves (FCV-LW-104A and FCV-LH-101,1\\).

b.

These monitors (RM-CC-105 & RN-C-C-106) automatically closes surge tank vent valve (IICV-CC-100).

c.

These monitors (RM-GW-101 & RM-GW-102) automaticnlly closes discharge from containment vacuum ~iys t_c111); and waste gas decay tank valves FCV-GW-160, FCV-GW-260, FCV-GW-101).

d.

These monitors (RM-RMS-159 & RH-RMS-160 or RM-RMS...;259 & RM-RHS-260) trip affected unit's purr,e supply and exhaust fans, closes affected unit's purge air butterfly valves (MOV-VS-100 A,B,C,&D or MOV-VS-200 A,B,C,&D). e

e.

These monitors (RM-RMS-162 & RM-RMS-262) trip affected unit's purge supply and exhaust fonr;, clone :1Cfecti*d unit's purge air butterfly valves (MOV-VS.. lOOA,B,C,&D or MOV-VS-200 A,B,C,&D).

f. These monitors (RM-SV-111 & RM-SV-211) divert flow to the containment of the affected unit by opcnl 11g 1'V-SV-102 and closing TV-SV-103 or opening TV-SV-202 and closing TV-SV-203.

ACTION 1 ACTION 2 3 SE.E. Cot.l..1,\\.t~1 4,e. ACTION 4 ACTI°ON 5 ACTION 6 TABLE 3. 7-5 (Continued) TABLE 1-"JTATION ..: * : -- -* j e With the number of channels operable less than required by the Miniraum Channels Operable requirer.:.ent, effluent releases mayA--b~sumad for up to 14 days, provided that prior to initiating a release:

1.

At least two independent samples are analyzed in accordance with Specification 4.9.A.l.e, and ~e

2.

At least two technically qualified members of the Station Staff independently verify the release rate calculations and discharge valving; otherwise, suspend release of radioactive effluents via this pathway.. ,o With the numbers on channels operable less~an required by the Minimum Channels Operable requirement,}~ffluent releases via this pathway may continue for up to M days provided that at least once per 8 hours grab samples are collected and_A,,(be.-'co.- analyzed for gross radioactivity..(at a J:.owe:r,. limit of detection of at least 10-7 µCi/ml. ~a-."1W..a-'i-BtTt.-op-i-e-a-na-:tj<-sJ:.s.- With the number of channels operable less than required by the Minimum Channels Operable requirement, effluent releases via the affected pathway may continue for up.to 14 days pro-vided the gross radioactivity level is recorded at least once per -8' hours during actual releases. ~4 ~ With the number of channels operable less~an required by the Minimum Channels Operable requirements~~=luent releases via this pathway may continue for up to~ aays provided the flow rate is estimated at least once per 4 hours. 30 With the number of channels operableG:ss than required by the Minimum Channels Operable requirement':),.effluent releases via this pathway may continue for up to -W days provided grab sam-ples are taken at least once per 8 hours and these samples are analyzed for gross activity ~-:i::-ga.mma-,i.s&tOfH.--c-ana-rys4:& within 24 hours. 3,1) With the number of channels operablec'i"~ss than required by the Minimum Channels Operable requirement':)effluent releases via this pathway may continue for up to~ days, provided samples are continuously collected with auxiliary sampling equipment..for -pB-HOO-S-HOt~e-a-t.e-~)-days. r-e..i l.l.l'.,.-ed. t "- -r ol,le, 4, '1-3.

--+>1:>Lf. 3.'1-S (CoyJi >\\.u.~) e f)GT101-~ 7... With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for gross radi91ctivity (beta or gamma) at a limit of detection of at least 10 microcuries/gram:

a.
b.

At least once per 8 hours when the specific activity of - the secondary coolant is greater than 0.01 microcuries/gram DOSE EQUIVALENT I-131. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 micro-curies/gram DOSE EQUIVALENT I-131. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that prior to initiating the release:

a.

At least two independent samples of the tank 1 s contents are analyzed, *and

b.

At least two technically qualified members of the Facility* Staff independently verify the release rate calculations and discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

        • 1
':'1

~J ~:-:-:-

  • .:*.1

- *-- ----------1***

e e TS 3.11-1 3.11 EFFLUENT RELEASE Appi°icabili ty: Applies to the controlled release of radioactive liquids and gases from the station. Objective: s To establish conditions by which gaAeous and liquid waste containing radioactive materials may be released, and to assure that all such releases are within the concentrations specified in 10 CFR 20. In addition, 'to assure that the releases of liquid and gaseous radioac- .tive wastes to unrestricted areas are as low as reasonably achievable as set forth in Appendix I to 10 CFR 50. Specification A. Liquid Effluents

1.

Concentration

a.

The concentration of radioactive materials released in liquid waste effluents to unrestricted areas from all J c;:t t..U.. ti l.}..2...~ l reactors at the siteAshall not exceed the values speci-fied in 10 CFR 20, Appendix B, Table II, Column 2, for disso[,,z.d_ ov-.uir~

  • radionuclides other than;\\noble gases,-a.nd-t-F-i-t-i-um-,,_or~d-

-no-t--r-esul-t:--:i-n-an-annua..1-a'l.te.r..age-eeneen-t:r--at-ion-o-f-t-r-i.t-i-Um- -OF-an-a-nnu-a-l-ave-r-&--6e-~-0nee-n.t-r-a.t-i-0n-o:f-di-sso-lve<l-nob-l-e-gas

5

-i-n-t-he--disGha.r,~e-ai1-a-l-i-n--e-xeess-of-4-x-10--pe+/-fnrl:.- ~ov '.11.:S~Dlvd._ o..- -w.-""C-~*w.. VL.OY.e. ~a..se.s 1 Tue.. C.0-,\\c.e_.._~- sk.u_ t?e._ ~.,,\\~ fb 2.x.io J,l-U.)µ_L L"ta..L o..c:t(v1+j.

~* 3 3.11-2

b.

1-:ith the conce,:t:-.2.~i.o.n of :r-adioactive :::-.aterial released from the site to u~restricted areas exceeding the limits to of A.l.a iwr::!ediate:y restore concentration/\\within the

2.

Dose +' C: Q.I O.I E .c: +' +' +' C *r-a, a, E UOl::lE

J C: 0- 0 Q.I "'O,,- a, u s.... >,c:

.c: Q) s.... Vl f'CI +'S....:::I..CQJ"'Of'CI -a ::J V1 C: OlO VlOQ.10 C: +' V1 -a,- +> +' a, rn "OCC:..CS....UE OJQJO.l+'O a, Q)~:::, s.... s.... U rn,- O'l OJ ::J E X+I-CVlO Cl/ 4-,,- 0 'I- 0 Q)QJS....-0 r-s....0

i QJ 0

-0 "'O Q.I I.fl -0 4-0 *,- > a, C: +' :::, -0 *r-.c: rel O"C+J+J V) V,,,- n::I l't:l -..... C:,- r-Ol -0 QJO S...::JC:O I.fl *r-c: a, E *r-..o

, +J.,.... +' :::, s....

rtlU S....U::Jr-UrtlV1rt1 -Ortl r-

i a,

+J QJ a, rtl 0-.C: Vl 0 .c: >,,- +' Q) +' +' *,- s... s... I.fl +' Q) n:, +' rtl a, Ill u +' "'O rtl Q).c: QJ a, l't:l C.C:r-+J s.... E C!J +' OJ 4-S.... s....o 0 Cl/ n:, 0 +' +'U>UU'l.C: c: U E Q) a, +> +'

J Q)

"'O.CUCVIV>S....

  • ,- +' CO a, S....

E. o s.... a, E .C: Vl *r-S.... +' 0 M UQJ"'O::lS....S....

C.~ ~ u ~ 4-.~ i
3.

3 4- <lJ O"r-.C:j aJ 4-.C rt! +' I +' -0 0 +J S.... ::J *r-1 S.... rt! °O ~I o -o v, 4- -o.,.... I Q. C Cl/ 0 C: > Vli QJ ct1 V) a., *r-,,- ~ ~ co S....r--0 l .......,Q)QJl'OC:Vl" V1,- -a u,,- s.... '\\ ri:,-.....a, C: Q). ,,- +' S....,,- Cl/ C +J C: u,,- l'O OJ rt! s.... rtl l a, E a, E s.... rO oil a.*.-.c: Q)..c: a :::, s.... I V),-+J S....+'+' O"O

a.

The dose or dose commitraent. per reactor to an individual from radioactive materials in liquid effluents released to unrestricted areas (See Figure 3.11.1) shall be limited during any calendar quarter to< 1.5 mrern to the total body and to.:£} mrem to any organ, ~ cJ..u.y; '4 ~- G;i..l~a.v-1-jeAV io .f. 0 MY"<?..IIA. 1 0 -+kL fcTu.l loocf.:i ~ +c ~ JO ~Al'~M_ ~ ~ OY'lJO.,-,

b.

With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the limits of A.2.a, prepare and submit to the/\\~~ission within-3& d-ays, purs~-t-e--Spec-i-f..i~~i-r~ Liquid Waste Treatment '* all -t LAA/!.$' Tr'2.?-.tMe.DJ: ()P£(2.,'7uf. TtA.e...

a. A :fhe ;Liquid -R-a-dwasteA~nge %ystem shall be/\\,used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected dose due to liquid effluent releases to unrestricted area~ (See Figure 3.11.1) o.of:.

when averaged over 31 days would exceed/\\~ mrem to the 0, 7-. total body orA.~ mrern to any organ.

c:.,. -g ~ .J .s C \\,. -:1- ,,) r-C, C-J,.., i.., c; ~ s 1,,1 J ~ cD ~ -~ J w - ...9 SJ rJ) 1 d

, c,:,

~ -: ">/ M B. -i:l. - c:a <<'l W ii!,,._ -rk L_D~ rcd..wO.'.i\\C.. tre.&1.t..,~ s't :;"tevA.. i~p~L~ ~ ~.,>..oy~ HeJt ~s: or e

t. /\\ '/ith :-ac:ioactive liquid,;.,*aste being discr,2.rged '\\-:ith:.iut treat2~nt and in excess of the limits of A.3.a, prepare i'v"L l (e..u... t9f C',l'J 0t~..:.. v r-ero and sub;:*.it to thel\\G£HRm:i:ss4-on,within 30 days,/\\*fH.. i.:r-suan*t---t:-G Sp'°~

.Spee-i-a.-ea-t:-hm-6--.. o~ a !ffti-r-ty-Da.y-l*!r-i-t-tenAReport,;,;hich includes the following information:

1) Identification of

-ti-'-.. i "-0 f 4' c...b l..

  • I\\ equipment or subsystems ~

and the reason for in-operability, 2) Action(s) taken to restore the inoperable equipment to operable status, and 3) Summary description 0-

  • of action(s) taken to preventArecurrence
  • Gaseous Effluents
1.
2.

Dose Rate

a.

The instantan:eetls dose ratet\\HT unrestricted areas (See Fig-ure 3.11.1) -tiue to rad-:ieae-t+/-ve-ma-t:-eFi-a-l-9--re+/-ea~

O...t ail_ i-, ry\\.e.S ) *

~a-seous.-e-f.f.+/-uent:-s-£-r-om the site)\\shall be limited to the following values:

1) the dose rate limit for all noble gases shall be~ 500 mrem/yr to the total body and~ 3000 mrem/yr to any skin, and 2) the dose rate limit for all radioiodines and for all radioactive materials in particu-late form and radionuclides other than noble gases with half lives greater than 8 days shall be~ 1500 mrem/yr to any organ *
b.

With the dose rate(s) exceeding the limits of B.1.a, WiHu:V\\., immediately decrease the release rate to -eemf>-i..-y-w:i:-t:h)\\the 6-i-Z,,f.tv, limit ( s ); ..a-H<l--p.r-O-V-:i:<l-e-~~ot..f-f..ica-ti:on-E-e-t-he-omm-i-ss-:i:on-

c.

Dose, Noble Gases 1 o:t ol.L ti Mes 1

a.

The air dose per reactor in unrestricted areas/\\(See Figure 3:11.1) due to noble gases released in gaseous effluents -~**

s.. 4-It! OU C S.. S..'QJ 0 ~ ~ ~ 4-, CU U "O

  • r-C It!

rc1 Q) Q) s.. EOJr-E QJ ~ Ctf s....cu..-.. +> C) <I.I S.. N _c.µ::,...._, +> C 0 Q) 4- -0 0)::, C CO"Qlrc1

  • r-

<I.I VJ r...VJOJC

J..O..C 0

-0 ::, +> *r-VJ +> VJ °' ro +J QJ C *r-C..C *r- -0 <I.I +> r... n:l

, r...

r-0)-0 4-C n, 4- *r-a, E a, s.. VJ E

, o ro VJ "O -0 0) 0 -0 (I) r...

QJ C > 0 VJ ro.,_ 4-ro +>

0) r... n:l -0 (I),-

C'CI C+J:JS..

  • r-s.. E E Vl:JU..-..

(I) 0-C) VJ QJ r- "' s....c...._,

0) C'CI +>

-0 C I C.µ *r-1 a., ro..c I r-..C.j.J C'CI.µ *r-C U 3: 0 0

  • r-

+> VJ V) +> C

  • r-rt1 I

~ VJ~ vi '.a r... r... s.. rO

, (I) (I) s..

u +> +> r... s.. res a, res res +> .c::,::,a, +J 0.0 I

b.

e 3.11-4 shall be limited~to 2._ 5 mrad for gamma radiation and~ 10 mrad for beta radiation during any calendar quarter,~ ~ 10 r,tro.d.. for.50.w.~ v-ru:L.o+/-:oV\\.. O.\\--.L ~ 2.D M.ro.cl. fcv k:id-JL.. l"o..dJ..o..t i D"- clwr i ""-"- ~ ec.J.e.t-..d--0-.V' l-1.eo.l", With the calcula~ed air dose from the release of radio-active materials in gaseous effluents exceeding any of the limits of B.2.a, prepare and submit to the~ A l;<... lie.v... et~ o-l-wz.v-vv.,O'('t re..tr.1ireLl b'1 Speu.fice..:lio... b,/o within 30 days,A.put=-&U-an.t-+/-0-Sp.ec-i-fJ.~.a.t.ion~, a 1 .Sp.e.~ ~i-r~-4-H-efi;,leport which identifies the cause(s) for exceeding the limit:§)and defines the corrective actions~ __.---- to 6a._ -\\ oJe. ~ +o r-u:lv...u.. -ti~ n.. l~ e.s c:+ y-o._cl~ GI...C.-'ti v e.. V\\.o k, l2....

3.

Dose, Radioiod~nes, Radioattive Material in Particulate Form, and Radionuclides other.than Noble Gases ) or cill. f '>.A.e..5 I

a.

The dose per reactor to an individual/\\from radioiodines, radioactive materials in particulate form, and radionuclides (other than noble gases)with half-lives greater than 8 days in gaseous effluents Figure 3.11.1) shall released to unrestricted areas (See 'TO CllVj O V' i o-"" be limited to 2,. 7.5 mrem/\\auring any calendar quarter, a,....d_ f IS eo...Lwca.v ~ ~.

b.

With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides w1'+J.... ~tf-_Hns 5v~u- +lNi..-. ~k( ~s 1 (other than noble gases~in gaseous effluents exceeding the OJA s+ -+~ J iYI. lit.v.. et M_j o+Lur re.rod r.ev..V:.rw.. ~ Sf12.t.t~i~l:.c,V\\. (!I,6, limits of B.3.a re are an'd submit to the...,-Gemm-i-ssion-within "Dtrc.c.-tor of tv~ Of+tc.e. q-L,..r,p~c.'c,o..-.. 14.0, ~erc.~w-u-:t ~; l)v-.c-.L 30 days, ~aR-t-t-o-S-pe~-i-ea-t:-:wri-6-;-6. 2. b, a ~-f=)---1}a-y-Spe_~ .I\\.;..Wr:i:-tt-en-Rep or t which identifies the cause(s) for exceeding Y"£cl..l/,.LI!...- tl...c.. re..L~ e.s c9i-- (" 1 1o - the limit and defines th~ corrective actionsJ:c b~ -to.ke.-.- +a/\\ iodines and radioactive materials in particulate form, and radio-nuclides (other than nobles gases) with half-lives greater than 8 days in gaseous effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose or dose commitment to an individual from such releases during these four calendar quarters is within (15) rnrern to any organ.\\

_ =-1-s

4.

Gaseous Radwaste Treatment

5.

'j s:te...._ Tr-'?;>--t~ ~l +l,..e.. V ev:c i lic.:-i o.-- b ~.:w.t Tr~.:t cJZvS. I\\ - a. The Gaseous Rad,,;asteA~-t-a-aen-Syste::i,\\shall be ~-to. OP£~ lt..c. Tt'\\11. OffYq)i";o.tL portic;1.s or fk, 60..RollS (kt~rte..1v-eai:""-u---\\ Lph....._ -s~ be.. u.i:,;,.cl t-o reduce radioactive materials in gaseous w~stes Prior to Ba~= l f-tr tu. e....X CW.- '; as ec u s their discharge when the projectedAdoses due toAeffluent wi-thout treatment and in excess of the limits of B.4.a J l v... Li e.u_ e-\\- 0-lN-j t:,-\\{..u' 'r"for-'i V-7\\ UJ.Y.iu.l lo'1 $pe..C:.:..+ i C$---l:, OJ.,\\_ (,,. i,, J prepare and submit to th_g}'... GomHi-iss4ea within 30 days, G:°1)iredo, o-\\, -H,.e. 0-H-ic.e.. of Tusptc.bov-a.,,.J.. E.1'.fer-c.e,,~el"-t ~ i o~ t9{+ I c.c.- ..pm:s~e *.c * -'f*h-:h-ey-D~itten Cl. Spec.u,..L AReport. which includes the following information:

1) Identi-

~ ~i,..of:(!J,'t>.Jok.. OV" fication of/\\equipment -&f'A.subsystems ~~~ and the reason for inoperability; 2) Action(s) taken to restore the inoperable equipment to operable status*; and 3) Surmnary description of action(s) taken to prevent a recurrence. Gas Storage Tanks

a.

The quantity of radioactivity contained in each g2s storage tank shall be limited to~ 26,700 curies of noble gases (considered as Xe-133).

e 3.:l-[;

b.

With the quantity of*radioactive material in any 6as storage, tank exceeding the limit of B.5.a, iITL."Tiediately suspend all additions of radioactive material to the tank and within 48 hours~~ reduce the tank contents to within the limit.~v4e-e-prompt notification te-t-he

  • ..f-0-J:.l.oW=up.-...r.epo.r-t-shall incl ude-a-<lescri-p-t-fon-ef-a.e-t:-iv it ie s
S-ase~.

2::::: Specifications 3.11.A.l and 3.11.B.1 are provided to ensur that radioactive materials released in liquid and gaseous e fluents from the site *:~nrestricted areas will not e~ceed t;h~ncentration / / limits specified-in 10 CFR 20, Appendix /able II. In liquid effluents, the co~ion limit ~oble gases is based on the assumption that Xe-135 is~e c~~rolling radionuclide and its }fPC in air (submersion) was conv~ed to an equivalent concentration in water using the method;;,A~ribed International Commission on Radiological Protec~n (ICRP) Publicatron 2. For gaseous effluents, the specif/;;;;;::l dose limits provide ~able assurance that indiv/,s in unrestricted areas will not be exp~ual average concentrations in excess of the limits specified in ~ix ~ble II of 10 CFR 20.

e e TS 3,tJ-7 3.n.').J B:l:i:s £pecification,._Js provided to ensure that the concentration of radio-active materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to an individual, and (2) the limits of 10 CFR 20. 106(e) to the population.* The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

3. lJ.r,.Z.:

E+i=s- ~pecificationAis provided to implement the requirements of Sections II.A, III.~ and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept 11 as low as is reasonably achievable. 11 _ The dose calculations**;n the ODCM implement {he requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, ~uch that the actual exposure of an individual through appro-priate pathways is unlikely to be substantially underestimated. The equations specified in the OOCM for calculating the doses due to the actua.l release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, 11Calculation of Annual Doses, to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,U Revision 1, October 1977 and Regulatory Guide 1.113, 11 Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I,U Apri 1 1977. 3,Il,~,2.- .:Ef::i:r:s..§pecification applies to the release of liquid effluents from each* reactor at"'the site. ~r units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system. The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept 11 as low as is reasonably achievable 11 Ih:i:s gpecification;Jmplements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

TS ~.il-6 e e 3.1/. B, I -Ih--is-§pecification10s provided to ensure that the dose at any time at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with*the concentration5 of 10 CFR Part 20,' Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20. 106(b)). For individuals who may at times be within the site boundary, the occupancy of the _individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to less than or equal to 1500 mrem/ year for the nearest cow to the plant. 3,IJ.1"$,1 "'ff:R::5 ~pecificationp~pplies to the release of gaseous effluents from all reactors al the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. 3.11, B. '2-- =thi=s ?.pecificationAis provided to implement the requirements of Sections II.B,.' III.A and lV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The dose calculations establi~hed in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1. 109, 11Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with l O CFR Part 50, Appendix I,U Revis ion 1, October 1977 and Regulatory Guide 1. 111, 11Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Rel eases from Light-Water Cooled Reactors, 11 Revision 1, July 1977. The ODCM equations provided for determining the air doses at the site boundary are based upon the historical average atmospheric conditions. 3.1[.B,, "!- ihis- ~pecificationAis provided to implement the requirements of Sections II.C, III.A and 1V.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Opera-tion are the guides set forth in Section II.C of Appendix I. The ACTION state-ments prqvide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept 11 as low as is

i. -**- -

-,_::I

      • 1 j

e e reasonably achievable. 11 The-ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcula-tional procedures based on-models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, 11Calculation of Annual Doses to Man from Routine Reieases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, 11 Revision l, October 1977 and Regulatory Guide 1. 111, 11Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled

Reactors, 11 Revision 1, July 1977.

These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The felease rate specifications for radioiodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which were examined in the development of these calculations were:

1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consump-tion of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILA-TION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems r - **--.. be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept 11as low as is reasonably _ achievable 11 * ~ ~pecificationl\\.implements the requirements of 10 CFR Par~3.l/.£,.4-50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given *in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.Band II.C of Appendix I, 10 CFR Part 50, for gaseous effluents. Spe~-f; ~t"-: 3, I/, £V5 1 b_j I\\Restrict.Tn-g the quantity of radioactivity contained in each gas storage tan~provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7. l, "Waste Gas System Failure 11 * . I ---**1 Li

1

.:::::1

e "iS 3.il-#, /0 I Fig. 3.11.1 Site plan of Surry Power Station.

4. S-1 e
4. 9 EFFLUENT SA~*fPLING.AKD RADIATION :t-m:nT0?..2::-;G SYSTE~-1 Applicability:

Applies to the periodic monitoring and recording of radioactive effluents. Objective: To ensure that radioactive releases are maintained as low as practicable and within the limits set forth in 10 CFR 20 and Appendix I to 10 CFR 50. Specification A. Liquid Effluents

1.

Concentration b, The liqttid-e-f--Auent e-0n-t-i-nuous-mon-i-t-or-s-ha-v-ing-p-rov-i-sions-:- -for-aut-0ma.t..ic-t-e-Fmi-na-t;...iBn--o.f--l-iqu-iii-releases-, -as-1-:i:s-t-ed-:in- -rad-+/-oa-ct+/-ve--mat-er-ia-l-r-eleased-a-t-any-t:-ime-f-rom--t-he--s-ite-t -un-1:est-r-:k-ted-a-r-eaS-to--t:-he--vaJ:.ueS-g-i-ven-4n-Spe--i.f..i~at-i-on.

c.

The radioactivity content of each batch.of radioactive liquid waste to be discharged s~all be determined prior to release by sampling and analysis in accordance with Table 4.9-2. Tke.. Y"esv..l-t-s o+ prt...- r.e...l~-e... ~ se.s sl.,..J..l k vs=l 1.>0i-h-.... -h,.e... ~ticw:tl_ rv\\e:\\i,-e:;l, l1>- ~ DDC...'-'. to o..s:s~ -tt..a...t +N._ c..m-..cL....:r;-'V:l.1':..o-* ().._t -t£...-.e.... po.Lv...( e-t Y".e..l~e_.. LS ~~-ro.i.:s.---u:1.. (_uj~ ii-L k.l.'-;rS t>f Sfe_q;u~ 3.L\\,11, I.O-..,

-rtv__ v~e..,:v..a... o-t -- pl'w: t>1,H po~t - r~e... 0-ws.lr,e.~- S[.,..-J..J... !.oe.. us...,\\.. w:-ti,.. --ft-p_ ~~~ M-e.11~ ;,.._ t'"-L, ODU/\\. +o 0-SSU.Ye- -n-..,...c" ~ C,(:S1}.~ oJ;; -H.....0- f 0°1ir-![" 0t l'l'..~Q..- We.t""- IV~ w)+h-,.."....._ ~ k..t.., .L Sf~ ' \\ ' '

  • j I\\ " I e I CLA..71 ~;-
  • .,;.,
  • a....
2.
3.
d.

Post-release analyses of samples fron batch releases shall be performed in accordance with Table 4.9-2.

e.

The radioactivity concentrations of liquids discharged from continuous release points shall be detennined by collection and analysis of samples iri accordance with Table 4.9-2. li'l(... r.e.su...Lh Eit ~ ~ses s~ lo'1..- os2d with. -tk e.:..lc..v..l~l 1V-e...ti-.c,d..t i:._ ~ ODD.\\. +o ~W"'<!.-, ~ -f1....e. Cci\\.- C~...-r.itcv-s o.t -H,..e... po\\lA., ~+ ire.L~e... (are... r~ w,f!....t'......., fr,..__Q.... Luv..,-K sf-Spe..ci..fic..4;.oy;_ 3, ll, A, I.~, Dose Dose Calculations Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calcula-tion Manual (ODCM) at least once per 31 days. Liquid Waste Treatment

a.
b.

Doses due to liquid releases -t:-e-uB-1=-e-s-trieted areas-shall be projected at least once per 31 days, iv\\. D.CC..ord..o..v..c~ w:+1,...., +i-. e... 0 p Uv\\, tv,..e0-.t. MUt The tiquid jadwastel\\~n-E-x-e:ha-nge iystem.shall be demon-strated operable/\\at least once per 92 days unless the liquid radwaste system has been utilized to process radio-active liquid effluents during the previous 92 days. B. Gaseous Effluents . 1. Dose Rate

2.

-b..--T.he-noole-g-as effl-l:len-E-ee-nti-nuous-morr:i:-t:ors-havi-ng-p-rc-- -v4.--&ions-f.o.r-t;-he-aul:oma-t--:i:-c-term-ina-t-:i.--on--of-g-as-eous-re.l,a-s-2-s-$' -as-1-isted-+/-n-T-ab:l:e 3-. 7 5-shaH-he-l:lsed-t-o--1-i-m-i-t....o-f-fsite-doses-w+/-thi~he values es t-ab:l-+/-s*hed-in-Spec-i-f-i-e-a-t:-ien 3, 11. B. r; xep-r-e-se-nt-at-:i:-ve-s-amples--aad-pe-r-f-&rIIri:ng=:ana-~es in accordance tJitk -fl~ Mc:t'"-Ods ~ pro*u.J.tAA..U of ~ ODLJ-t. -w-i~the-s-runp-ling-and-aaa-l-y-s-i-s--p-FB-g-r-am,-&!}ee-H-ied in Tab+/-e--

d.

The dose rate -4:a-Hnrestrieted----areas.., due to radioactive ~ateri- _als,other than noble gases}~-leae-ed, in gaseous effluents~ of- ~pe.C..:.f-lCA--f:,'ei"'-- l-, 11, E, l. tv shall be determined to be within the..-.=eqtri-r limits,\\by----u-tifi§ Dose, Noble Gases Dose Calculations~'

  • Cumulative dose contributions from noble gases in gaseous r +'

l t , Tor *~ C-ufr u.<X u. ~ i LA-Cv" w a.,-A C,,U,-r~1 u:.le-i....lOY '1 ~,--, effluents/\\shall be determined in accordance with the Offsite Dose Calculation Manual (ODCH) at least once every 31 days.

'11 2 LlJ C 'f-Q.) f-V)

r V) V) >-

OJ...-

  • ;::)':::, V)..C. 'I-

<( 0 +' '1-

c LlJ 1-,

(l) XV) :Z: V) LlJ <C LlJ Vl U') {!) :E: QJ ::;

z:

f-,- 0 OOJ<CCOJ 1----i..c. LlJ :J U') f- +J o:: ro c::c f-U') C) -' 0) 1----i C f-11:l <V f- *,- V) 'U >

z: +J :::,

LlJ 11:l<CN+' >s...::r::cnu OJ X 11:l 'U O.LU S... O CO QJ*,- 11:l z

0. 'U

>,o n:,

E:..0 1----i QJ s...

LU f-U f-LU <C C Vl V)-' _J O l/) >- a:,....... QJ V)c:Cf-+'U a::: Z Vl 0 f-LU UJ 11:l S... Zc...> OJ 0.. LUO ~ 'U 0 f-'U C+'+l <( QJ 11:l ltl LU +l 'U a::: ltl+J ~a, f-S...CVJN +l QJ QJ *,- LU Vl E +l..- f-C 0. :::J *,- V') 0 *,- C +J <C E :J *,- :::r 3:0JO'"E Clt!Q) Cv, ~QJ::::E ,~~ V)..0 ~ ..0 ""O

,,- V) +' V) C'J 0..- >-

Vl ltl O"l UJ ltl V) ltl..c. V)..c. QJ U') <(v,f-,-E:::r (.!J z: QJ 0

E: w +l +' *,-

QJ UJ :::E; rc, VJ> ..c. f-f- >, OJ f-V)c:CS...VlS... >-WO

a.

V)* a::: 'I-QJ f- +J a, I I. ~ UJ t.~ :5 1* LU I-QJ S...

E:V'lEO.CJ f-c:CO.OC

<( 3 *,- s... *,- f-0::'.QJltlU 3, Do -- n 1-' 1 P

, e,

!",2.C i c, i o o J.. :1 es, :-.=. =:. o ~ c_t_i_*,_, e_._._a_t_e_r_i_.::._* ____ 1._n __ a_r_t_1_c_u_l_f:_; t_e __ F_o_::.--_,"...,_* 2:1d Ratic~uclides O:~e= T~a~ ~oble Gases Dose Calculations --41' +k eu....... e.... t. i~v ~~ _c.u.Yv~"t c.c,...le.,.d..Cl.V LJ=.1.. Cumulative dose con tribu tionsAf.Fom-r-ad4--0nue-l-i-des-C't--her-t:han- .. nob-1-e-gase-s-in-g-aseotts-e.f.f.J... uen-E-S* shall be determined in accord-ance with the ODCM at least once every 31 days. 4; Gaseous Radwaste Treatment

5.

~. Doses due to gaseous releases to unrestricted areas shall be projected at least once per 31 days, '1v... G-C.c.ev~ w;t[_ -tke.. l9DC.M,*

b.

Gas Storage Tanks The quantity of radioactive material contained in each gas storage tank shall be determined to be within the limit ,0... established in Specification 3.11. B. 5Aa t least -<ttJ&-rterly-, EltA.c..e... pev-24-br:I ~k~ v-c..d.loo..aive... ~-twto..l:. o.n. !oe.i~ o..cl..o-w.. -to the.. "t"c1.iA.k., '5el',A.l

a:,~ g.~3 J
c.

Reports Thexinnual Radioactive Effluent Release Report shall inc;:lude the informa_tion specified in Section 6. 6. 3. b. ~~-"---* ---------------

.----~~-~~

/ ~ o.v-.d.. Y"t>.d.i.oo.c.tirL <1c..sc.ou.i -~.J-.f-Lv..e.,...t D{' Each radioactive liquid effluentAmonitoring instrumentation channel ~ shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE / CHECK CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table~ ..... 4,1-1. ~- - - --- ---- ---.---*. -*~ E. The environmental program given in Table 4.9-1 shall be conducted. F. A census of animals producing milk for human consumption shall be conducted at the beginning and at the middle of the grazing season to determine their location and number with respect to the site.

TS

4. ? The census shall be cc~ducted under the following conditio~s:
1.

Witl-.in a 1-mile radius from the pl2nt site or within the 15 mrem/yr isodose line, whichever is larger, enumerated by a door-to-door or equivalent counting technique.

2.

Within a 5-mile radius for cows and a 15-mile radius for goats, enumerated by using referenced information from county agricul-tural agents or other reliable sources. If it is learned from this census that animals are present at a loca-tion which yields a calculated thyroid dose greater than from previously sampled animals, the new location shall be added to the surveillance program as sooh as practicable. The sampling location having the lowest calculated dose may then be dropped from the surveillance pro-gram at the end of the grazing season during which the census was con-ducted. Also, any location from which milk can no longer be obtained may be dropped from the surveillance program after notifying the NRC in writing that samples from milk-producing animals are no longer available at that location. Bases The specifications contained in this section establish adequate sur-veillance requirements to ensure that the limits of Section 3.11 of the Technical Specifications are not exceeded. -T-he--d0.se-e-a--lcul-at-:i:ons-contained-in-t-he--ODGM-imp-lement-t::-he-Fequi-rernen-t-s--- o-f-Seet-ion-I-ll-rA-oL.Appendi-x-I-to-lO-CER.-SD..;-t.hat con f o..mance-W-i-th--t:-he -g-u-i-<les-o-f-Append:h.-+-:i:s--t:-e--be-shown-by-ealcu+/-a-t-.ional-me-t:-hods-based--on--.

e

4. 9-6

£,'ppropr:f:-at-e-patb-.*ays is unlIKely t8 be substant:ially und::::resti~ I mated. The equ~tio~;;~c~~ied i~ the ODC*1 for ~-tf!ating the I

... ~

doses due to the actual release rates 01:-><r-adioactive materials

I.

in liquid and gas~~are consistent witli'-tbe methodology pJovided i:_ Reg,tlatory Guide 1.109, Revison 1 and inco~ gi~.E.EG:::ill.-1--3-, Octob-er-I9-7-~ The test and calibration requirements are specified to detect possible equipment failures and to show that maximum permissible release rates are not exceeded. All the~t~o~ -the--a-bo-ve-sp-ee-i-aed--t:es-t-a-n<l-ea-+/-l:-i>Fa-t-i-on-f--r-eque-n&ies-a-re--adeq-ua-t-e-.- The environmental survey incorporates measurements to provide back-ground data and measure possible plant effects. Samples collected at points where concentrations of effluents in the environment are expected to be the greatest will be compared with samples collected concurrently at points expected to be essentially unaffected by station effluents. The latter samples will provide background measurements as a basis for distinguishing significant radioactivity introduced* into the environment by the operation of the station from that due to nuclear detonations and other sources.

e e 4.S-7 This schedule *,..-i.11 er;.sure that changes in the c!Viro:.1.-:7.ental radioactivity can be detected. The materials *which first sho-;., changes in radioactivity are sampled most frequently. Those which are less affected by transient changes but show long-term accumulations are sampled less frequently.

s.u....d ~--'.)!-!.)..&,a,,\\,: ~, ,6 ~ 14 e T.L.3L:. !..,

  • 9-2 RADIOACTIVE LIQUID WASTE SAl{PLING lL~1) A.'\\ALYSIS PROGRR-f Sampling Li "d

.QU1 Release Type FrequencY* A. Batch Waste Re-p lease Tanksd Each Batch

l. li4 ~

QoJ..wc.r*h_ £~w.wfs p One Batch/M p Each Batch p Each Batch B. Plant Continuous -w= Releasese M Grab Sample

i. S'teruv... G.o...~

&. ow d..eu.1 "'- [

).. T v-.v-1.,l~ -g~

'i) '()v-~ Grruo ~k w Grab Sample Iv\\ G rob So.ur le.. Minimum Analysis FrequencY* p Each Batch M M Composite b Q Composite b M ~ w ~os,-1-e. I-M C . t omposite Q Compositef Type of Activity Analvsis Principal Gamma Emit terse I-131 Dissolved and Entrained Gases H-3 Gross alpha P-:,2.. Sr-89, Sr-90 Fe. - 5S Dissolved and Entrained Gases Principal Gamma EmittersC I-131 H-3 P-":>2-Gross alpha Sr-89, Sr-90 Fe.-SS I Lower Li::7J.it of Detection (LLu) (µCi/ml)a 5 X 10-7 b 1 X 10-6 1 X 10-5 1 X 10-5 1 X 10-7 I ';(. 1n-b 5 X 10-8 l)(lO-b 1 X 10-5 5 X 10-7 b 1 X 10-6 1 X 10-5 l ~,o-6 1 X 10-7 5 X 10-8 Ix,o-b

e T.!.. BLE 4. 9-2 ( Cc:-:~i:1ued) Tl-J3LE,-.-.~-. --.-..,., .;,_,;._: J.---.!..;...U._\\

  • FREQUENCY NOTATIO~~

w M Q p At L e.iut OV\\.C..1:.- pe-r 24-lttoLJ.Y-.S 1ieckJ:)"- Prt l-eo..~t 01-\\..C..cz... f-eA.- 1 ~

s.

Mo:-. th!y-A+: l ~ t O"N:...e.. p A.. 3 I ~ ~ -Qtiarter+/-y* Pit: l~t 0-.--C,.L-r~ '1'2- ~ s Completed prior to each release

a.

The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement sy_stem (which may include ra.diochemical separation): LLD= E *. V

  • 2.22 x 106
  • Y
  • exp(-).!it) where ll CL pr1on LLD is theAlower limit of detection as defined above (as µCi per unit mass or voltnne)

Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute) Eis the counting efficiency (as counts per transformation) Vis the saople size (in units of mass or volume)

2. 22 x 106 is the number of transfo_rmations per minute per microcurie Y is the fractional radiochemical yield (when applicable)
t.

( r e e

, is the rc.c
:.:2::ti.ve cecay constant for the particular radionuclide r,s.id...po;~ of 6t is the elapsed ti~e between/\\sample collection {o~-end.-of th-~

o.~** -S.::.=p-:1-e-£;:;.l.1-e.c-tion~dr-atAtime of counting. The value of so used in the calculation of the LLD for a detecticn system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In ealeulating-t-he -P-t:ese.nt-4.-n-t-he-&amp:l-es--Ee.g.-,po-t-ass4:um-4-0-4n-mi-1.-k-sampl-es}-.-:- T ;1 p,* c:,;J_ voll.ce..s of E.11/1 'i, a....J... lt s k.LL b(.... vs:u:l \\,.__ ~ ~ ~ -+/-t-&hGu.-ld-be-Feeegn-iz.e<l-t:h~he-LT:,D-.is-def..:i:ned-as-an-a...,>r-i:o-r-i-(-bef-o-r-e-t-he- -f.ac-t-}-1-iai-t-~r-6efrt4ng-the-e-apab-i-1-i-t-y-of-a-meas u rcement-sys-t-em-and-not-as..

b.

A composite sample is one in which the quantity of liquid sampled ist--f-e.r-e.-e.r~ proportional to the quantity of liquid waste discharged and in which ' the method of sampling employed results in a specimen which is a4:enue.J:-i-d.e.s* representative of the 1 i.qui ds re 1 eased. '"'*...i-U::~o'"ns-n-ea-r-t:!-ie-tl:.o;:::::::under-4hes~s-t-ances,-t-he---LLD-nay-be ~ase<l-iw.re-r-s-e-l.y-p-roper-t:-:i:-ona-l~o-t-he-magn-i-eu4e---e-f-t-he-g-amma-y+/-cl:d- {-i.....-e.,- ~x~e-re-!-is-tli~t--0-n-abundan-ee-ex-p*ressed-as-a-<leclaa+/-- -f--'F-ae-t-4on-},--e-ttt-i-rr-no-ease-shal--1-the--L-I:;D,-as-ctleuhl:*t:-ed-4:n-t-h-is-mann~~r-a

c.

The principal gamma eraitters for which the LLD specification will apply are exclusively the following radionuclides: }fn-54, Fe-59, Co-58, Co-60, Zn-65, Ho-99, Cs-134, Cs-137, Ce-141, Ce-144. This list does not

d.
e.

'< : ** -.\\'.1 e ~~an that only these nuclides are to be detected and reported. Other p::E..}:s \\*~h.ich are :::ec:.su.r2.ble and identifiable, together with the above nuclides, shall also be identified and reported. ~~tte:l-i-des-wlri:eh-fr~e-* -aet-:i:ve-E~luent-Reieas e Repo-rt:-r A batch release is the discharge of liquid wastes of a discrete volume. friov-To sruu..pt;.l ~ ~ses 1.e.o,.c)..... b.:,..tc.l--.. sk.U.. kie.... is:ob.t"""w...., a,,,..,)._~ ~lj 0"1 ().... M.e..+kocl O..esc.,,lki.ul l"'- ~01)C.M 1 -n:, Clf;Sl,I..V<L- ~pVl'.Se"'-.t't;..t/ve.... :,.a.AAflt"'"'e), A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume or system that has an input flow during the continuous release. ~ .r.~* 0-'i> SfeL'-i To be_repr:sen!at~ve of the quantities and concentrations/of r~dioactive materials 1n ]1qu1d effluents, samples shall be collecte~~t,nuously -i-n~ fl ow of the efTl-ueftt-s:t~e:am-. Prior to analyses, all. samples taken for the composite shall be throughly mixed in order for the composite sample to be representative of the effluent release.

Gaseous Release Type I A *. Ventilation Vent System

n.

Process Vent System

c.

Waste Gas Storage Tank D. Condenser Air Ejector E. A~l Rel~a}e Types as listed in -/r&'B-above. A,i,ai-.ct.D . / (~"... - -- ~ TABLE 4.9-3 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Sampling Frequency*

W* JY\\b,c.,e..

Grab Sample =W-N\\ b,c.Je.. Grab Sample p Each Tank Grab Sample M b,c.,a.. Grab Sample Continuous; Continuous J'..

Co"'-\\, ~uo~~ t Continuous.f-Cot.Ltir-lJO\\J.S f I

Min:l.mum Analysis Frequencv* w: IY\\ b ~ =W-M. b p Each Tank Mb wd.. Charcoal Sample wcA. Pr1rticulate Sample ~ ~M G:.Mfo~ih. Pc-.v'ticc,..brt. . So-11.1.f l e-Q Composite Particulate Sample l'-.lobte.. fiat tv\\.C> ~ tc,y, j Type of Activity Analvsis Principal Gamma Emitters9 H-3.. Principal Gamma Emitters9 H-3 Principal Gamma Emitters9

H
;;;:3:.

Principal Gamma Emitters9 H-3 I-131 I-133 Principal Gamma EmittersC Gross Alpha SR-89, Sr-90 l\\olole.. GitUe~* Gvoss Bcit;... ~ 6c:..IW!ll.c-. Lower L:l.mi t or Detection, (LLD) C I ) µCi ml a 1 X 10-'1b 1 X 10-6 1 X 10-l1b I I 1 X J.0-G - 1 X 10-l1h ) +/-::=x=-to+/- I 1 X 10'.:""4b 1 X 10-6 1 1 X 10-12 1 X 10-10 1 X 10-ll I 1 X 10-11 l 1 X 10-l J. I I -b l )('. ID ,... -*- - - ---* L I r.

e TABLE 4.9-3 (Continued) TABLE NOTATION

  • FF~QUENCY !~OTATION w

M Q p -w~ At lea.st O;,,..c..e... pe,,r -z ~

s.

-Menthl-y-At le.tut" e,.._ u... p ev $ l ~ s -Qt1are-e-i=-~ At L~t' ~~ p w- ~ 2... ~ s Completed prior to each release e

a.

The lower limit of detection (LLD) is defined in Table Nota'tion a. of Table 4.9-2. H:rD-m-ay--be-4-nereased-4..-R¥e-:1.;sel~or-<Eionally to-t:he-ma~e- .gamma--y:i:eld-{4-.-e-.-r-:l:-x-:1--0::;::,,4/r, wheI;e-I-i-S-t-he--phe-t-on-abu-ndance-expressed see... i 'I t) T'N""O~

'..l \\,*', v\\.e K.. t p9e.

~-" Th 1 f h" h h LLD "f"

  • .u

':flpl:u e principa gamma emitters or w ic t e speci ication~111 app+/-yl\\. ~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-58, Co-60, Zn-65, No-99, Cs-134, Cs-137", Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. ~~vh+/-ch .ar-e-be-1:ow----t!Te-tLD--fur=--t-he-ana-lyses--s-heu-l<l-no t-b-e--repo-r-t-ed-as--beiag-p-FeS-en-t-

e b, Analyses shall also be performed following shutdown, startup, or a THERM.AL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period. c.. Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded. cl. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing (or after removal from sampler). Sampling and analyses shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour.- When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10.

e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

t* The ratio of-the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications '3.tl.l?,,l,6l.J :::..n.e.,.2.a..., 0-,i...d__ 3 I l \\. (!,-, 3, 0... o

Kote: e e r2dic2c~i~E ~aterial resulting from the fission p=ocess. Sealed sources or calibration sources are not included under this item. Leakage of*valve packing br gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.

3.

Unique Reporting Requirements

a.

_In-service Inspection Evaluation. Special summary technical report shall be submitted to the Director of Reactor Licensing, Office of Nuc~ear Reactor Regu"iation, NRC, Washington, D. C. 20555, after five (5) years of operation. This report shall include an evaluation of the results of the in-service inspection program and will be reviewed in light of the technology available at that time. 7f";-

b. SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT--:S':--S-S--~~?~-~-'

of the within ~~e1~l)i-ei;n-t:9?..- D .L,..,,-"' C, Routine radioactive effluent release reports covering the operation unit during the previous 6 months of operation shall be submitted 60 day~ -~fte_r January 1 and July 1 of each year. ~ The radioactive effluent release reports shall include a sum~ary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, 11 Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radio-active Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, 11 Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. The radioactive effluent release report to be submitted 60 days after January 1 of each year shall *include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, and atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation -:f A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. Amendment No. 14 II I

  • I I I

doses due to the radioactive 1iauid and ga5e~us effluents released from the unit or station during the prev{ous calend1r year. This same report shall also include an assessment of the radiation doses from radioactive liquid and I gaseous effluents to members of the public.due to their activities inside the site_ boundary (Fi gures/.-5-:-=l 3 ~nd-5.--1-=-41 during the report period. A 11 as sump-~ 3. 11 1 tions used i_n making these assessments (i.e., specific activity, exposure tir;ie and location) shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for deter-mining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM). The radioactive effluent rele~se report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power OperatiAi. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1. The radioactive effluents release shall include the following information for each type of solid waste shipped offsite during the report period:

a.

Container volume,

b.

Total curie quantity (specifiy whether determined by measurement or estimate),

c.

Principal radionuclides (specify whether determined by measurement or estimate), 8-Fa-a--Ofl.ttn.#R~i-m>,m noh I~ ---

d.

Type of waste (e.g., s~ent resin, ccrr,pacted dry waste, evaporator bottoms),

e.

Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

f.

Solidifieation agent (e.g., cement, urea formaldehyde). The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis. The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) rr.ade during the reporting period.

Arnendr7:ent No. 14 --7-76 2~radfoao-ti¥e.-!last~Summarize,LM~ntshcly)--Data-slr~. '. \\ of each year.

  • I

\\ (a) To~ount of solid waste package~ cubic feet). \\ (b) Estimated t:. tal radioactivity in curies)* involved. \\ j (c) Dates of shipmen (3)1 Fuel Shipments - In~ i and spent fuel shall be r (a) Date o~ments and d" position (if shipped off-site) \\ io~tive to each shipment of new\\ provided,"--including the following: \\ \\ / \\ (b) N~ber of elements shipped ?enHHcati:arr-ntnube; and ennclmfent of ~ e~i}}

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