ML19339C093
| ML19339C093 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 10/01/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| NUDOCS 8011170564 | |
| Download: ML19339C093 (77) | |
Text
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(o UNITED STATES g
NUCLEAR REGULATORY COMMISSION g
WASHING TON, 0. C. 20655 5
j October 1,1980 Docket No. 50-409 Mr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601
Dear Mr. Linjar:
By letter dated August 14, 1979, you proposed to ame nd the existing Technical Specifications of the La Crosse Boiling Water Reactor, for the radiological effluent and environmental monitoring systems, to implement the provisions of Appendix I to 10 CFR Part 50.
Our review of the proposed Radiological Effluent Technical Specifications was based on the model Radiological Eftluent Technical Specifications for Boiling Water Reactors, NUREG-0473, Revision 2, July 1979.
Our comments and a marked-up copy of your proposed radiological efflu-ent Technical Specifications are contained in Enclosures 1 and 2 to this letter. Yoe should incorporate these changes into your resubmittal.
Your proposed ame ndment did not include the required Technical Specifi-cations on solid radioactive waste, system operability, curie content in outdoor liquid holdup tanks, noble gas release rate, and administrative controls. The enclosed comments and marked-up specifications do not include comments on your Offsite Dose Calculation Manual.
To date, we have not received this submittal.
It should also be noted that the proposed amendment references Specifications 3.0.3 and 3.0.4 of the Standard Technical Specifications (STS), which are presently not part of the LACBWR Technical Specifications. Counter parts to BWR STS 3.0.3, 3.0.4 and accompanying 3.0.5 should be proposed and submitted to the NRC for review. An example of the 3.0.3 through 3.0.5 specifications and a Section 6.0 are enclosed for your information.
8011170
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Mr. Frank Linder. October 1, 1980 You have not submitted a Process Control Program (PCP) for solidifica-tion of radioactive wastes for La Crosse Boiling Water Reactor, nor referenced the PCP in your solidifcation system' specifications. Whether you use a contractor for waste solidification / dewatering or perform your own waste processing, a PCP should be submitted. We request
-that the PCP be submitted for our review, and that a response to the
-enclosed comments be made within thirty days of receipt of this letter.
Sincerely, L a k. L A h u) den'nTs' M.' Crutchfield", C@lef '
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Operating Reactors Branch #5 Division of Licensing
Enclosures:
1.
NRC Comments 2.
Marked-up RETS 3.
Sample Specifications 4.
Sample STS Section 6.0 cc w/ enclosures:
See next page 4
h A
Mr. Frank Linder October 1, 1980 cc w/ enclosures:
Fritz Schubert, Esquire.
Director, Technical Assessment Staff Attorney Division Dairyland Power Cooperative Office of Radiation Programs 2615 East Avenue South (AW-459)
La Crosse, Wisconsin 54601 U. S. Environmental Protection i
Agency
- 0. S. Heistand, Jr., Esquire Crystal Mall #2 Morgan, Lewis & Bockius Arlington, Virginia 20460 1800 M Street, N. W.
Washington, D. C.
20036 U. S. Environmental Protection Agency Mr. R. E. Shimshak Federal Activities Branch La Crosse Boiling Water Reactor Region V Office Dairyland Power Cooperative ATTN: EIS COORDINATOR P. O. Box 135 230 South Dearborn Street G:noa, Wisconsin 54632 Chicago, Illinois 60604 Coulee Region Energy Coalition Charles Bechhoefer, Esq., Chairman ATTN: George R. Nygaard Atomic Safety and Licensing Board P. O. Box 1583 U. S. Nuclear Regulatory Comission La Crosse,. Wisconsin 54601 Washington, D. C.
20555 La Crosse Public Library Dr. George C. Anderson 800 Main Street Department. of Oceanography La Crosse, Wisconsin 54601 University of Washington Seattle, Washington 98195 Mrs. Ellen Sabelko Society Against Nuclear Energy Mr. Ralph S. Decker 929 Cameron Trail Route 4, Box 1900 Eau Claire, Wisconsin 54701 Cambridge, Maryland 21613 Town Chairman Dr. Lawrence R. Quarles Town of Genoa Kendal at Longwood, Apt. 51 Route 1 Kenneth Square, Pennsylvania 19348 Genoa, Wisconsin 54632 Thomas S. Moore Chairman, Public Service Comission Atomic Safety and Licensing Appeal Board of Wisconsin U. S. Nuclear Regulatory Comission Hill Farms State Office Building Washington, 'O. C.
20555 Madison, Wisconsin 53702 Ms. Anne K. Morse Alan S. Rosenthal, Esq., Chairman Coulee Region Energy Coalition Atomic Safety and Licensing Appeal Board Post Office ' Box 1583 U. S. Nuclear Regulatory Comission Lacrosse, Wisconsin 54601 Washington, D. C.
20555 U. S. Nuclear Regulatory Comission Mr. Frederick Milton Olsen, III Resident Inspectors Office 609 North-llth Street Rural Route #1, Box 225 Lacrosse, Wisconsin Genoa, Wisconsin 54632
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' A -
P COMMENTS ON LA CROSSE BOILING WATER REACTOR RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS) 1._
We have reviewed the subject. radiological effluent Technical Specifications as submitted by the licensee, 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, contents, and table. format, to make them conform more closely to the contents of NUREG-0473, Revision 2, in an attempt to streamline these tpecifications. Specific changes made may require subsequent discussion.
2.
In Section 1.0, add definitions 1.33 through 1.37 and modify definitions 1.30 and 1.32, as shown in markup.
3.
Modify the following Specifications as shown in the markup:
3.3.3.8 4.11.1.3.1 3.11.2.3 3.3.3.9 4.11.1.3.2 4.11.2.3.1 3.11.1.1 3.11.2.1 3.11.2.4 4.11.1.1.1 4.11.2.1.3 4.11.2.4.2 4.11.1.1.2 4.11.2.1.4 3.11.2.5 4.11.1.1.3 3.11.2.2 4.11.2.5.1 3.11.1.2 4.11.2.2.1 3.11.2.6 3.11.1.3
- 4.
Delete the following' Specifications as shown in the markup:
4.3.3.8.1 4.11.1.1.4 4.11.2.1.5 4.3.3.8.3 4.11.1.2.2 4.11.2.2.2 4.3.3.9.1 4.11.2.1.1 4.11.2.3.2 4.3.3.9.3 4.11.2.1.2-4.11.2.4.1 4.11.2.5.2
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5.
Add Specifications for the following:
i Liquid holdup canks maximum curie contenc (Specification 3.11.1.4 a.
of NUREG-0473).
b.
Noble gas release rate-(Specification 3.11.2.7 of NUREG-0473).
i Ventilation exhaust treatment system operability (Specification c.
3.11.2.5 of NUREG-0473).
6.
--Modify Tab'es 3.3-11 and 4.3-11 as shown in the markup and address the a
following:
Gross radioactivity monitors on the liquid radwaste effluent line a.
should provide automatic termination of releases.
b.
A gross radioactivity monitor should be provided for the component 4
coolinI water system effluent line.
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c.
A. flow rate measurement device should be provided for the discharge i
canal.
d.
List the tank le.el indicating devices.for any outdoor tanks poten-tially containing radioactive liquids in accordance with NUREG-0473.
In Table 4.3-11, is the turbine condenser cooling water line monitor-i e.
'the same as the service-water system effluent line monitor listed i
in' Table 3.3-117 i
I 7.-
Modify Tables 3.3-12 and 4.3-12.as shown in the markup and-address the following:
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i The offgas. storage vault discharge monitor should have capability.
a.
to alarm and automatical.ly terminate release.
I b.
.The offgas treatment system explosive gas monitoring system should have redunJant hydrogen monitors with automatic control functions to reduce the potential of a hydrogen explosion. Acceptable types-of control functions are delineated in NUREG-75/087, Standard Review Plan 11.3, Revision 1.
^
c.
. Particulate. activity monitors, items 1.b, 3.b, and 5.b of Table s 3.3-12 and 4.3-12, are not required to be listed per NUREG-0473.
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d..
The containment building and stack effluent release points should have flow rate measurement devices.
In Table 4.3-12, expand on Table Notation item (3).
e.
8.
In Table 4.11-1 add sampling for P-32 and Fe-55, and modify Table Notation as shown in the markup.
9.
-Revise Figure 3.11 '.. This figure shall consist of.a map of the site area showing the perimeter of the site and locating the points where-liquid. effluent leaves the site.
If on-site water areas containing radio-active wastes are utilized by the public for recreational or other-
-purposes, the points of release to these water areas shall be identified.
The figure shall be sufficiently detailed to allow identification of structures near the release points and areas within the site boundary.
where ground and surface water is accessible by members of the: general i
i public. 1See NUREG-0133 for additional-guidance.
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- 10. Modify Table 4.11-2 as shown in the markup in accordance with NUREG-0473.
'11.
Modify solidification ~ system Specifications 3.11.3.1, 4.11.3.1.1, and a
4.11.3.1.2 as shown in markup to include the Process Control Program.
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- 12. Figure 5.1-1 is referenced in a number of specifications, but has not been
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prsvided.for review..This' figure shal l consist of a map of'the site area I
showing the-perimeter'of the site and locating the points where gaseous effluents are released.- If on-site land areas subject to. radioactive j
l materials in gaseous waste are utilized.by the public for recreational or-I other purposes, then these areas shall be identified by occupancy-factors and the-licensee's method of occupancy control. The. figure shall be suf--
I ficiently detailed to. allow identification of structures and release point-elevations, and areas within the site boundary that are accessible by-members of the general public. See NUREG-0133 for additional guidance.
I
- 13. Base 3/4.11.2.6 discusses explosive gas-mixtures 'and automatic control features, however, based on our review (see also comment 7.b), there are no tutomatic. control features in this regard.
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- 14. Modify BASES Section as shown in the markup.
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- 15.. Reporting requirements should be spec.fied in the administrative controls section of the Technical Specifications.
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. 16.. In general, the ODCM is not~ acceptable in'its present. form. Guidance for preparation of this document is contained in'NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants."
j Some specific deficiencies.are-as follows:
No description provided of effluent monitors.and their corresponding
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a.
i alarms and trip setpoints.
Instruments should be specifically
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identified.
i b.
The setpoint methodology discussion should be expanded to explain how actual setpoints.for each device are arrived at. This should include factors such-as instrument sensitivity, instrument error, effects of other' effluent streams on the setpoints, and the conserva-j i
tisms incorporated. Sample'setpoint calculations should be provided
.for each device.
Credit for in-plant dilution flow in establishing monitor setpoints c.
is allarable only.if sufficient safeguards exist to terminate re-leases upon loss or reduction of dilution flow; i.e., interlocks with dilution flow sensing devices.
i d.
The methodology employed for establishing gaseous effluent monitor setpoints appears cumbersome and >should be reconsidered in light of Section 5.1.1 of NOREG-0133.
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L ENCLOSURE 1 PROPOF"O CHANGES TO TECHNICAL SPECIFICATIONS t
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4 1.0 DEFINITIONS
- CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as nec',sary, of the channel output such tnat it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.
The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.
The CHANNEL CALIBRATION may be performed by any series of sequential, over-lapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.5 A CHANNEL CHECK 3
.11 oe the qualitative assessment of channel be-havior during operation by observation. This determination shall include,-
where possible, comparison of the channel indication and/or status with other indications _and/or status derived from independent instrumentation channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.
b.
Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
SOURCE CHECK 1.29 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
0FFSITE DOSE CALCULATION MANUAL (ODCM) 1.30 An 0FFSITE DOSE CALCULATION MANUAL (00CM) shall be a manual contain-ing the methodolnay and parameters to be used in the calculation of offsite doses due to rao.vactive gaseous and liquid effluents and in the calcula-tion of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints.
" equi,mmm..t: cf the 00C". m a r. usided !- Sp;;!fic; tion 5.!E.
LACBWR l-1 T
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o 1.0 DEFINITIONS - (Cont'd)
GASEOUS RADWASTE TREATMENT SYSTEM 1.31 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 i
or holdup for the purpose of reducing the total radioattivity prior to release to the environment.
l VENTILATION EXHAUST TREATMENT SYSTEM 1.32 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in parti-culate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulates -fr the gaseous exhaust stream prior to the release to the environment uch a system is not considered to have any effect on noble gas effluent J. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT i
SYSTEM components.
DOSE EQUIVALENT I-131 f.35 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."
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PR0 CESS CONTROL PROGRAM (PCP) l.54 The PROCESS CONTROL PROGRAM shall cont 6in the sampling, analysis, and formulation determination by which SOLIDIFICATiDN of radioactive wastes from liquid systems is assured.
SOLIDIFICATION ~ N.-
136 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).
i LACBWR 1-2
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l' PURGE - PURGING
- l. 3 t -
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.
VENTING 4 37 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.
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J TABLE 1.2 FREQUENCY NOTATION NOTATION FREQUENCY.
S At least once par 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
1 0
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
M At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days.
R At least once per 18 months.
I S/U Prior to each reactor startup.
P Completed Prior to each. release.
N.A.
Not applicable.
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LACBWR-1-3
.i INSTRUMENTATION liloMITORa dG RADI0 ACTIVE LIQUID EFFLUENT, INSTRUMENTATION
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LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation t"
channels shown in Table 3.3-11 shall be OPERABLE with their alarm / trip setpoint'sbithi" the spri#ied "~itc to ensure that the limits of Specification 3.11.1.1 are not exceeded. Tke. alar =/ trip setehts of Rese. cL= t.
T2Me ?.? . gsae be CatcMat;ou ha (opca).
alfk. -ike. o SLALL be cle.termiet iw accorclaue APPLICABILITY:
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At d -tim,
With a radioactive liquid effluent monitoring instrumentation
~ ACTION:
+
a.channelalarm/tripsetpointlessconservativethanph;<;iuc-chc.;n in T; Lie 3.3 11 *ich cn;urc; that the !"it c f 2. ' i.1.1 -
-C 2 mct, immediately suspend the release of radioactive liquid f.Cy
-c r:
effluents monitored by the affected channel or declare the Ti
&P-channel inoperable, o~
F&
less f4 % die wher effluent monitoring instrumentation channels %oactive liquid f"f b.
With en; ei m;;cof th; ;bovc iqu: es radi g
cpcr;b!c, take MM the ACTION required by Table 3.3-11.
PT The provisions of Specifications 3.0.3 and 3.0.4 are not c.
applicable.
SURVEILLANCE RE'}UIREMENTS 4.3.3.0.
T;.c me; gun.;> auli L: dete minmi in eccccda.ce with precedurc; c'; dc.ic. ibed in the CC " and chal! bc rcccrtd on th; Licuid L';';t: Ectch Icm L 31.
4.3.3.8.2 Each radioactive liquid effluent monitoring instrumentation channel snall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANflEL FUNCTIONAL TEST, and CHANflEL CALIBRATION operations du-4 ; the cordition; and at the frequencies shown in Table 4.3-11.
4.3.3.3.3 "ccre
%ditable reccrt-ch:11 bc m;intain;d, ii ouviden:c-with pr;=dures in +he CDC", of all radic2:ti'/c li';uid ef#luent "snite. %
in';trumen ta tica d a-/ trip ';c tp;i n t';.
Octpoint and :etpcint celculativ.w ch ll bc ;v;il;b!c for re'fiew to ensure th t the "-it';
cquircd br
-Specificaticii J. !!.1. ' crc mct.
LACBWR 3/4 3-44 t
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T ABLE 3.3-11 r-
'N RADI0 ACTIVE LIQUID EFFLUEhi MONITORING INSTRUMENTATION E"
MINIMUM CHANNELS APPLICAGLE
^Lf""/TPJF INSTRUMENT CPERABLE CCMC:T!ONt GCTTGINT ACTION I.
Gross Radioactivity Monitors Not-I Providing Automatic Termination of Release st I
a.
Liquid Radwaste Effluent
%gt Line 1
"' "11 Tir s CCC". "
51 46
'b.
- Service Water System Effluent j
Line 1
^t ^11 Ti.,~a OC0" !
52 f
2.
Flow Rate Measurement Devices a.
Liquid Radwaste Effluent Line 1
at
^'l Timcr 44MEF 53
^
TABLE 3.3 (Cont'd)
TABLE NOTATION
- ACTION 51 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements.. effluent releases may be resumed for up to 14 days, provided that prior to initiating a release:
1.
At least 'two independent samples are analyzed in accordance with Specification 4.11.1.1.3, and; 2.
At -least two technically qualified members of the Facility Staff independentyi verify the release rate calculations and discharge valving; ll a Otherwise, suspend release of radioactive efflui cts via Cits' pathway.
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ACTION 52 With the numbers of channels OPERABLE less than required by the Minimum Channels OPERABLE requiremg,nt, effluent releases via this. pathway may continue for up trfpb4 days provided that at least oncer per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> grab samples are collected and. analyzed for gross radioactivity (beta or gamra) at a
' lower limit of detection of at least 10-7 uCi/ml.
ACTION 53 With the number of channels OPERABLE.less than' required by the Minimum Channels OPERABLE requiremg_nt, effluent releases via' this pathway may continue for up tqv$ days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during 'the actual releases. Poup cuves q be., usut 4e e.st;uro p.
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LACBWR 3/4 3-46 "hMMh@
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TABLE 4.3-11
,RADI0 ACTIVE LIQUID EFFL'UENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS s
CHANNEL CHANNEL SOURCE FUNCTIONAL CHANNEL INSTRUMENT CHECK CHECK TEST CALIBRATIQ 1.
Gross Beta.or Gama Radioactivity Monitors Providing Alarm But Not Providing Automatic Isolation l'
~~
a.
Liquid Radwaste Effluent Line' D,# W JPP
-Q($).
R(4)
" # b. a..
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.surbine Condenser Cooling b.
E6 Water Line D
M Q(1)
R(4) g a
w 2.
Flow Rate Measurement Devices a
a.
Liquid Radwaste Effluent Line D(3)
N.A.
Q R
TABLE 4.3-11 (Cont'd)
TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
1.
Instrument indicates measured levels above the alarm / trip setpoint.
2.
Circuit failure.
3.
Instrument indicates a downscale failure.
A. htrued Cutrols not set in CPERAT6 wAt.
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- 42) A CnmaEL CnECs 2 [oiUni.eiS2ti"ft'y?" h a "" "I"
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.c..; hour of bc;f W g of re'enre (3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNCL CHECK shall be made at least once per 24 leurs
% on eny dad on which continuous, periodic, or batch releases are made.
(4) The CHANNEL CALIBRATIO.1 shall include the use of a known liquid radioactive source positioned in a reproducible geometry with respact to the sensor and emitting beta and ganina radiation with the fluences and energies in the ranges measured by the channel during normal operation.
J LAC 8WR 3/4 3-48
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ei INSTRUMENTATl_0N Moonob %
-a4 d
.y ij RADI0 ACTIVE GASE0US EFFLUENT, INSTRUMENTATION e s g o T+
<s LIMITING CONDITION FOR OPERATION
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The radioactive gaseous precc= rd effluent monitoring instru-set 3.3.3.9 M.t J' mentation channels shown in Table 3.3-12 shall be_0PERABLE with theirj d
alarm / trip setpoints witbi-the spec #4:2 12
- +
to ensure that the limits of Specification 3.11.2.1 are not exceeded. 'The(. h/-tr:p sap;.4 e 4
j g, a y d '2 ekamts sLatt. la hteretnel A accordane iML. h otcg,
s APPLICABILITY: As shown in Table 3.3-12.
n t
With a radioactive gaseous procc= cr effluent monitorin-
_o ACTION:
a.
yyy instrumentation channel alarm / trip setpoint less conservative 4y Tab!
3.3 12 ;SicF ca:urc: that the
.h' 5 ? h glen it.s~ h mietimum.wder of thanAthe -f i c thcu 4-24 F
l 'it: cf 2.11.2 ' 2rc ut, declare the channel inoperable.
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~j OF effluent monitoring instrumentation channels *peous prc;;Z b.
With c= cr nr^ radioactive ga 3
ir cr;ble, take the ACTION required by Table 3.3-12.
The provisions of Specifications 3.0.3 and 3.0.4 are not c.
applicable.
SURVEILLANCE REQUIREMENTS
- 4. 3. ?. 9, ' Thc 4 chit; shall bc detc H ned in accu.Januc
- u. U1 g r = = -e n dc=ribed ' ttc CZ: and ai oil be. %ccdcd 'n the clibratic r;;c, J3.
4.3.3.9.2 Each radinactive gaseous pr==
cr effluent monitoring instru-mentation channel shall be demonstrated OPERABLE by perfonnance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALI-BRATION operations &+; the c~ditic= =d at the frequencies shown in Table 4.3-12.
- 4. 2. 2. 0. 3 ^udit:ble ecordt sh? be maintained cf thc calcol; tion; Tade, u ;;=rdreco m.4 + h nrncga,,rge 4n the gcy, cf 3 r;dj;nti fc peceu-31 Md 0#"e"* ""'"4 t eri a; i n3"umcr.t ti c-2 7 2 r'"/tri p S cipGinis. 3c; Lpv i si ti cad ctpoint =l c!:tient : =11 bc evailable fsc cevicw to cn,urc that the l 'it; rep! red by Spccific;ticr. 3.'i.2.1 cre met.
l LACBWR 3/4 3-49 t
i I
E TABLE 3.3-12 9
RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE INSTRUMENT OPERABLE CONDITIONS
-- Aimn/TRIF 5ETF0 INT -
ACTION 1.
Main Condenser 0ffgas System Monitoring System (10-Minute Holdup Tank Effluent) a.
Noble Gas Activity Monitor i
49EM:F 58 see b.
System Flow Rate
,., g,e
-Measuring Device 1
_ ':^^
54
%,gt 2.
Offgas Treatment System m
. a, Explosive Gas Monitoring o
. System a.
Hydrogen Monitor q O Hydr;;cn By '!:h :
56 3.
Reactor Containment Building Ventila' !on Monitor System
^ D CM., -
55 a.
Gascous Activity Monitor 1
1 see b.
Particulate Activity N
55
%,,a Monitor 1
-A c c.
Sampler Flow Rate
{ ^'"%
54 Measurement Device 1
1 O-
TABLE 3.3 (Cont'd) r-N j
RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE INSTRUMENT OPERABLE CONDITIONS
^L?""/T",:r ';"T;G;;;T ACTION 4.
Offgas Storage Vault Discharge Monitor (After Treatment System) a.
Noble Gas Activity Monitor I
n n,,, -
gg
~ ~ - "
b.
System Flow Rate MONE--
54 Measuring Device 1
w2 5.
Stack Monitoring System w
o, a.
Gaseous Activity w 59 00 Eft t Monitor 1
[**
b.
Particulate Activity 55 nn ~
Monitor I-7.c.
c.
Iodine Sampler
,, n e_r,
'w. _ _
o momm n ' --
57 1
m m.
mm
.trid;c and Sy:ter Ope":ti0n
~~~
d.
Particulate Sampler y,,. A ^r, c m.. m m v '. " ". --_
57
' '~'-
1
,s_
r2u__
nunm,
6 a,.d Sy:te-Ope n tien -
e.
Sampler Flow Rate O- ~
54 Measuring Device 1
m
1 TABLE 3.3-12 (Cont'd)
TABLE NOTATION ff A ttys.
%,r:ng rc, ::::.:: th :. : p a th'.:;y.
- During offgas treatment system operation.
- hrN opea.h et h mk cnuicasa de ejector.
g ACTION 58 With the number of channels OPERABLE les; than required by F
J the Minimum Channels OPERABLE requirement, gases from the main condenser offgas system may be released to the T
nvironmentJprovided:
g
.e 1.
The offgas delay system is not bypassed; and Sier disr b,n e The nffgas dge va.dta..:y sy', tem noble gas activity monitor 2.
is OvERABLE; 12-Otherwise, be in at least HOT STANDBY within ile hours.
g ACTION 54 With the number of channels OPERABLE 1ess than required by the Minimum Channels OPERABLE requiremtnt, effluent releases fg
" [4 via this pathway may continue for up tq@8 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
s 3 R *L E r ;2
+4 ACTION 55 With the number of channels OPERABLE less than required by 9 f_,, [
the Minimum Channels OPERABLE requirement hef'!uent re!^ :::
y vi; this p;t."ay may conti:.u fcc up La 2v day; pr^;f ded grab e
t;Ecn-at icast on c per 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />, and thc c ::rple;
!!"? ct cr
<J d.
'c; onaljicd for grc;; acti 'ity withm 2 h o u rr.
5 P
V one.
ACTION 56 Witii the number of chcnnels OPERABLE less than required by A
the 'linimum Channels OPERABLE requirement, opegation of the j
offgas treatment system may continue for up toff & days. k) tk b q
(
previded rec = bin:r t=p;ceture: cre men'toccm h;urly and
,g g p.
,,,o.c..,_.,,
_m.,, u... u, _....
4.u.
3 Nh
'If ACTION 57 With the number of channels OPERABLE less than required by b
the Minimum Channels OPERABLE requirem
, effluent releases 5j' via this pathway may_ continue for up t days, provided y
samples a_re continuously collected wit auxiliary sampling Es l1 rd k
equipmentffer ;;ri;d; = t% ctr f==n (7} d:y 3"?!y cd 4 t"- 19 heur :fter th: cr.d-Of th:,=p? "; p^r W.
J
@ Tion 59 With ShI number of channels OPERABLE less than required by the
[7 Minimum Channels OPERABLE requirement, effluent releases via 7-this pathway may continue for up'to 30 days provided grab samples are taken at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and these samples LACBWR are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
g l
C.
4
~
TABLE 4.3-12 g-RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL SURVEILLANCE CHANNEL SOURCE FUNCTIONAL CHANNEL.
REQUIREMENT INSTRUMENT CHECK CHECK TEST CALIBRATION CONDITIONS l
1.
Main Condenser Offgas System Monitoring System (10-Minute Holda Tank Effluent) a.
Noble Gas Activity Monitor D
M Q(i)
R(3) V b.
System Flow. Rate Measuring a
M. ice D
N.A.
Q(7)
R
~7. c, 2.
0ftgas Treatment System Explosive Gas Monitoring System w
a.
Hydrogen Monitor D
N.A.
M Q(4) 3.
Reactor Lontainment Building Ventilation Monitoring System j
a.
. Gaseous Activity Monitor D
M Q(1)
R(3)or(5) see.
Q(1)/
R(5)t/
c... a b.
Particulate Activity Monitor D
M
- 1. c.
c.
Sampler Flow Rate Measuring Device D.
N.A.
Q(7)
R.
'I
j-l TABLE 4.3 (Cont'd) r-
'l M
RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATIO'i SURVEILLANCE REQUIREMENTS CHANNEL SURVEILLANCE.
CHANNEL SOURCE FUNCTIONAL CHANNEL REQUIREMENT INSTRlNENT CHECK CHECK TEST CALIBRATION CONDITIONS 4.
Offgas Storage Vault Discharge Monitor (After Treatment System) b),
f@PtWJ, R(3) u a.
Noble' Gas Activity Monitor D
M f
b.
System Flow Rate Measuring Device D
N.A.
4 A.Q R
5.
Stack Monitoring System j
a.
Gaseous Activity Monitor D
M 0(2) [
R(3)or-(5)
Y$
See b.
Particulate Activity
- d Monitor D
M 0(2 ) !-
R(5) /
7, c.
c.
Iodine Sartpler 27.. ';:
D.
N.A.
N.A.
N.A.
d.
Particulate Sampler J".t.
D N.A.
N.A.
N.A.
e.
Sampler Flow Rate Measuring Device D
N.A.
Q(7)
R L
.i
TABLE 4.3-12 (Cont'd)
TABLE NOTATION 4
,,0t M,_ V..Y _h __......y.u...,..
m,_
- m.....
- During offgas treatment system rc =M an operation.
WKat %f oprm,tios o{- %. uak. tenfesser giy ejeckog,
The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic (1) isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
1.
Instrument indicates measured levels above the alann/ trip setpoint.
2.
Circuit failure (provides con trol room annunciation alarm only).
3.
Instrut.ont indicates a downscale failure (provides control room annunciation alarm only).
- 4. Instrument (matrots ru>t set in operate. mde..
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following coneitions exist:
1.
Instrument indicates measured levels above the alarm / trip setpoint.
2.
Circuit failure.
3.
Instrument indicates a downscale failure.
- 4. Inst <went (wb>ti ut stt % opu=Te, udt.
(3) The CHANNEL CALIBRATION for radioactivity measurement instrumen-tation shall be perf ormed by analyzing the gaseous radioactive stream for specific activity.
(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1.
One volume percent hydrogen, balance nitrogen; and 2.
Four' volume percent hydrogen, balance nitrogen.
i LACBWR 3/4 3-55' 1
.-.._.,n.n.
n,-, n,_ _
~
~
i i
TABLE 4.3-12 l'oTt'6)~~
TABLE NOTATION The CHANNEL CALIBRATION shall include the use of a known radio-(5) active source positioned in a reproducible geometry with respect to the sensor and emitting beta and gama radiation with the fluences and energies in the ranges measured by the channel during normal operation.
The CHANNEL FUNCTIONAL TEST shall demonstrate that the control (6) room local alarm occurs if the instrument indicates measured levels above the alarm / trip setpoint.
The CHANNEL FUNCTION TEST shall demonstrate that the control room (7) local alarm occurs if the flow instrument indicates measured leve's below the minimum and/or above the maximum alarm / trip setpoint.
3/4 3-56 LACBWR f
2 t:
. c' w m e
.3]4.11_ RADI0 ACTIVE EFFLUENi_S, E
gg, g
3/4.11.1 LIQUID EFFLUENTS u5bI-EuE"E
%5"##
CONCENTRATION tcI55 03#.5 LIMITING CONDITION FOR OPEPATION vu% s. -
.M*33 3;5Ei 3.11.1.1 The concentration of radioactive material released at any time Er 8 5 o 5 from the site te U rcctricted cren- (see Figure 3.11-1) shall be limited
- >,0#5 to the concentrations specified in 10 CFR Part 20, Appendix B, Table II,
%" '55 Column 2, for radionuclides other than dissolved or entrained noble gases.
"IE8E For dissolved or entrained noble gases, the concentration shall be limited E "yg to 2 x 10-4 pCi/ml total activity.
n 3.:"
- 5;
,L 3 3 APPLICABILITY: At all times.
5
~.2 gjegg ACTION:
8 25'
>, 3 3'3 With the concentration or radioactive material released from the site-te 8
t t-d exceeding the above limits, immediately restore con-d5*-,
wm. c, c }
- u. mcentration within the above limits.=d pm;it pre..pt nutificctier to t%
3.5 5
t133g.
c a i=4r purre nt to sp =i<i= tica c..l. m.
8%eten E"E557 2 0,u 3 o ;
SURVEILLANCE REQUIREMENTS nNha*H c-u m
- 2UEE8 4.11.1.1.1 RC cr.ccrt'ctica of rcdi' nti'fe ctcrici t ry ti c i -
A c"LL liquid ;ff.uent5 relcred #rr th 'itc chcl! bc continu;a51, niun Lorcd
{3'523
" nccrec.= ith We?.?'
EdiUC j ~g; y 4.11.1.1.2 The radioactivity content of each batch of radioactive liquid 3,4 g g a waste to be discharged shall be determined prior to release by sampling and analysis in accordance with Table 4.11-1.
The results of pre-release analyses shall be used with the calculational methods in the ODCM 10 assure that the concentration at the point of release ish" mite 7 b %
7
'!M ux 4' Specifi cation 3.11.1.1.
7{
coucosited y{
a1 p
Post-release analyses of sampleshiram batch releases gds)*t-4 g' 4.11.1.1.3 The results of theg be performed in accordance pith Table 4.11-1.
li-release analyses shall b8(with the calculational methods inthe 00CM to assure that the concentrations at the point of releasepe '-ited t;-
4 E*
'- Specification 3.11.1.1.
d@E F
"^ "
y@pf
=.u w ts m4~ m=tm :fm.=mpm ill includc the u.fe. <.atics cp=i' icd in specifiuticr. C.0.1.3.
-?. f,.
5 F.
LACBWR 3/4 11-1 w eper
.ea=ura se
..-.o..e-a-m e4
.em.
..ee
o TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM
.ower Limit Sampling Minimum Type of Activity of Detection Liquid Release Type Frequen,y Analysis Analysis (LLD)
Frequency (uCi/ml)a P
P A. Batch Waste Re-Each Batch Each Batch Principal Gama 5 x 10-7
- 1 ease Tanksd Emitterse 1-131' 1 x 10-6 P
One Batch /M M
Dissolved and I x 10-5 Entrained Gases 1 x 10-5 P
H-3 Each Batch Compositec Gross a 1 x 10-7 I x so-C ~
- p. 37.
P Each Batch Q
Compositec Sr-89, Sr-90 5 x 10-8 Fe,- ss 1 x to '
LACBWR 3/4 11-2
' gr y,,. ~
. '..,, m _ w. y,.,.. y.,,g,,...g,. z.,
--m....---,
e
O D
D WD N3 i
n l bG Ogg TABLE 4.11 _i_
c (Cont'd)
TABLE NOTATION The lower limit of detection (LLD) is defined in Table Notation a.
a.
of Table 4.12-1 of Specification 4.12.1.1.
3.
r;r ccriai rad!cnuc!i t: "ith !c;. gr; yic!d cr ? ce er^rgier, y for certa!" -?dicruc14e 4;turc, it mcy not bc pc :ibic t
.c a u r e-4" concentration: car the LLD. "n&r th::: c i ca" r dicnu:!!&c ct>r% the LL:'..;j te increased 4 yerre!y prnnnr+4ca211y tc the "2;ricedc of um p.-; yicid
'i.e., 5 " 10 7/I, tccc I i; the pheten : Loc.Jon w e m ca,cd :: a dec4-'1 'r'ctien), but 'n 90 c2:e
? sreci'ic r:dic-LLO, as wiculeted in this = nner for ch2T' th:
nuclit, bc ;ccatcr th;n ICI of the.".FC isisc sp :ificd in 70 CI", CO, "ppendi' ?, Tabic !!, Column ?.
A composite sample is one in which the quantity of liquid sampled is c.
proportional to the quantity of liquid waste discharged and in which the method of sampling employed re.ults in a specimen which I, representative of the liquid released.
Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representa-tive of the ef fluent release.
d.
A batch release is the discharge of liquid wastes of a discrete volume. Prior to samptr for aAses, tuk. batcL sLau.be. isolatut, ad
-tkee neroo3 g miuA,
a we.%d clastrit.aA : h ootw, +o assm.
t representdin sampiq.
The principal gama emitters for which the LLD specification will e.
apply are exclusively the following radionuclides: fin-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144.
This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, toriether with the above nuclides, shall also be identified and reported. ".:lisc:, u.ich arc ic'-
th: LLO fcr thc ne!yrer. N net bc repc-tci re beim; pc ;cnt at th; LLC ica l. An unu s, :!
'rce t;a;;; cc uit 'n LLD'; high;r than rcquiiud, die ica cm chal' 5: tr.c:nted i,-thc scm,1c nu:! %dic ctivc Ef'!urnt ",cicant "c p ; r t.
t LACBWR 3/4 11-3
~ ~. _ - -
=.
~
l 43* 33'39" N. L AT.
8 y s "
S SCREEN g
HOUSE N
t
[
I k SCREEN HOUSE
\\'
,^s,
/
g' LACBWR
/\\
/
\\
\\
/
g
/
s
/
60" O!SCHARGE
/
CIRCULATING WATER q
's
/
\\
/
90" CIRCULATING WATER DISCHARGE-
/
DISCHARGE LINE STR.
's
/
/
J I
y- ]
/
H
/
I
(
GE!10A N0. 3 A
a PLAN l
SCALT F FEET F
I I
0 100 200 300 400 m
Figure 3.11-1 LACBWR 3/4 11-4
==* -
+ =.....
'f
A t
4-Y RADI0 ACTIVE EFFLUENTS d-DOSE _
E LIMITING CONDITION FOR OPiRATION i
bm h ch, 3.11.1.2 The dose or dose c.ormitment to an individual from radioactive s
'3[
materials in liquid effluen?.s released tc cr: strict d r:n (see 3
Figure 3.11-1) shall be limi ud:
a.
During any calendar quarter to 1 1.5 mrem to the total body hy and to 1 5 mrem to any organ, and L
b.
During any calendar year to 1 3 mrem to the total body and to
{
< 10 mrem to any organ.
E APPLICABILITY: At all times.
o T
ACTION:
l I-With the calculated dose from the release of radioactive 2 7 a.
,f 1 materials in liquid effluents exceeding any of the above limits.
Aprepare and submit to the Commission within 30 days, pursuant f
to Specificatinn 6.9.2, a Special Report which identifies the
{', /
cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during thg remainder of the h
l dr quarters so that tha"psc.;gd during thqtubsequent t ree ca en current calendar quLrgge dose or dose commitment to an individual from such releases during tFesa four calendar j
quarters is within 3 mrem to the total body and 10 mrem to any organ.
b.
The provisions of S%cifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.2.1 Dose Calculations.
Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calcu-lation Manual (00CM) at least once per 31 days.
ki Ef'
$ ud
- e t
LACBWR 3/4 11-5
RADIOf.:iiVE EFFLUENTS LIQUID WASTE TREATMENT g
1:
LIMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE. The j
system shall be used to reduce the radioactive materials in liquid wastes
+
prior _to_ their discharge when the projected dose due te liquid effluent L
releaseVt: re:*ricted ;.sa; (see Figure 3.11-1) when averaged over 31 days would exceed w mrem to the total body or 4,4 mrem to any organ.
0.06
- 0. z.
APPLICABILITY: At all times.
O
% k*i d M k " F # " 1"I" E h ACTION-a.fXith radioactive liquid waste being discharged without treatment and in excess of the above limits,nprepare and subn 't to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following informat1s 1:
S e i m ern v e-Identification of equipment or subsystems et O'EP2 ELE and 1.
A the reason for "e-epc M i " ty; ino pr M ith.
inojeud.1e.
2.
Action (s) taken to restore thegexg. ;Mc-equipment to OPERABLE status,a d 3.
Summary description of action (s) taken to prevent a recurrence.
b.
The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS l
I.
4.11.1.3.1 Doses to liquid releases t = ::tricted va r shall be a
projected at least once per 31 days, a acurh. astA % opcm.
The liquid radwaste system shall be demonstrated OPERABLE ht 4
4.11.1.3.2 least once per 92 days unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 da.vs.
wid. radeado 4rtAtme,I qdw-pee *I M af,
b3 o y al 4L4.
least
,*as,3 1
t 1
I l
LACBWR 3/4 11-6 l
~-
.1 l
LL O
RADI0 ACTIVE ffrLUENTS 3/_4.11.2 GASEOUS EFTLuff(TS, DOSE RATE
_L.IMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate et ;g ui..~
th urr:^tricted arc;; (::c-0
.5 riger: 5.1 t due to radioactive materials released in gaseous effluents m
from the site shall be limited to the following " '--
V 15
'l e;
5 (see F 0us. 5.r-s)
The dose rate limit for noble gases shall be s 500 mrem /yr to m
mae a.
the total body and.<. 3000 mrem / year to the skin, and O >ce co-r nca c m.-
The dose rate limit for all radiciodines and for all radioactive m
na 5 2 g.*
b.
materials in particulate form and radionuc!f des (other than noble
- o2 gases) with half lives greater than 8 days, shall be < 1500 s.
E.5 ) *
"5xg mrem / year to any organ.
-au
- 2 a s.
." n o8 APPLICABILITY: At all tines.
M D G.U
~-
c a 5 3 "..c.
ACT10N:
,w;g qt,_,1,ws h@,
a k
EY With the dose rate (s) exceeding the above limits, immediately decrease the i-3..2.'
-d p-.~idc 9 j 2 s,C release rate 6 g' ~ ;'!',
i th uiu i n.dL(2) giver 3 o S % e' prr;t notifkoi.;vn w; th: CC-i n f or. pu r:=nt to Sp :t 'ic ti co. 0.9.1.12.
uaec weoe.
Ono 6-o o E $ ".5 >*
SURVEILLANCE REOUIREMENTS s- -c2e u
o-eo-
_mt,_
- 2 y
,4.
ou ony m e v.
omm.m
.m u> e..u-
...e
, e ene, oce
.,. m. m m
,,, 4, g
+ u. ~.,. m... _
m-
_m m,
m
__..__,,_2 x,,
ww uvsw
. u ww ua c a 6GW 6 4 2 6.c u wwvew o a._
Ti.ws allu s s wu wwri wi vi
.wm j
ygg ocacu wav-rn ssri fi c a t 4 nn L11 n a ** e m e 5CJga 4.11.2.1.2 Thc n^hle ;2: cfuer+ contin"n"e ~0riter
';# g prc'vi;icr.
no a
- i, " a 5
'cr th c u t m a i.. u tc r-i r.:ti c= or "a r: u:;. eleasc;,
, listcd in Tivie -
em ite dneae withia t '.: ;-. lues a te'i+ed 8. 5 5 b,
-3. 0 _12, 2 Le ii be u- ",te limi t 3.
_,. c __;
.t_.,
m m_ m i,_,_,,.,
,_...._m, m
,a.
- v. m. m. _.12.
m...i....,.
x m
v.
c 2,- m o,
- 4. I 1. 2.1. 3[ " :
.n 5,5 $.;t re! n: -.t r ' re!! nt '..
" % h, c'h c then-e % +
c E 3 s, gator. ' ; ncu; cffiucat 5heii bc d;t:r"'n:? by cbtain'ng reprewatat M
- 55%
c: p'^
r p;rfor-4a; ara!yce:
- 7 nord=;c i.i th the sa,,pi',y 2nd '
1;;;!: -
d
- m. - _ i o_
,, 3 3 4.
___4., w. -
-u c
=
W U@ m r v,p.um, sy-.
o e s.
a av d S5E%
- 4.11.2.1. qA_ U C DOI 2 P3t
""'^I t P' C tC d 2'^2 E._ d". _ t o P3d I C" ti iC '
C'~ '
4
- ^
4
.u.,_.
.m_3. m_ s e m a 4,,
m..., w.r.
u..,,.,
a.
u.o
. m. w
/
-+u.
..vu M.3,,
3:_4.,wg wj
_r
_ti_
.uu,.,,
vm ut w
. ggw w.
.so e, my
_ _,.. uw,g
+
+
m.
..,4._.,y.
. t_
+ m. u. w u. u. c w.%
3_
gJ s
wg w s ig w
w
,,,m. 4 c 4. m_ a_ 4., r, u i. -....
o,,
1.._. ym. vi un o3 a,,
.t_
s.m.-.a++ ens
.-...s i m,_4, som
,A_)
wwJ.
,, n. n. e +, 4
---,,t,gJ-g_
- 2+a 4n
,2 m e m wc.
g,.
y g,bbJ E,,
B
'qi
^ ---
,a 4 4.. e o..,__.,
Tha w mi s n _ _ _ ',
n_,><
_, +<. - c r. e.,. _-., + o t.-
ac nuev.
, e m...<
A :.
n
.m_
c, _ _ _ 2. r t _,., + 4. m m.. c. n.
' n v '. u m m-.... 4 m ra m + 4 mm.
J sk.
a 'no 'n 'i
..s.
i c...
The dose rate due to noble gases in gaseous effluents shall be L
d;termined to be within the above limits ir accordance with the methods and pr:codurm, of the ODCM.
t
\\
s k
5-TABLE 4.11-2 9
!s RADI0 ACTIVE GASE0US W:dTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Sampling Inalysis Detection (LLD)
Gaseous P.elease Type Frequency Frequency Type of Ateivity Analysis (uCi/cc)a P
(utw bw
.P
.!Each h e.le Principal Ganma Emittersh
-1 x 10-4b
,ain Cr.i..;d Of f-NGi*
A. 4"Cn Sy t : ' O "' a utc Grab Sample Fqd "c!d; 5'- Cffhcat Ead-P q c.
4 -3 i x to 9 e
B 'Off-Gas Storage M
Each Sa:rple Principal Gamma Emittersh 1 x lo _4b Vault 3ischarge Grab Sample
/
3 C.
Stack Effluents Continuous 9 We C I-131 1 x 10-12 w
Charccal 7
Sarple I-133 1 x 10-10 fO co Continuous 9
':5 Principal Ganma Emittersh Particulate (I-l?,1, Others) 1 x 10-11 Sarple I
Continuous 9 M Spgi Particulate Gross Alpha 1 x 10-11
- Sample 1
i Continuous 9 i Q Cce;;csite iParticulate Sr-89, Sr-90 1 x 10-II i
lSu'ple i
i
, is.t. is:
Nobk 6ates 448 E;-;' L.., sic M
-1. 10 ' '
Gross Beta ed. 6auw l
i y so-b Cm&lsooos yg w l
igio 4
- y. Sta r E % <A 1 MI i
i Pn.
4 b S.:ttm
.up l
Ke g-s i g io _ s i
3 *a fe 8'T
' Gr. N I
M Lad h
,.1; e,,,
Mi 09
-.m_ e. c,.,
m.
.. <, o.n,.....,,...
.r u c, TABLE 4.11-2
'" f,j.O * ; " '" c '
~ Tcorit d)~
7-
.o me..%
V L L: m O
- t. O 3
C eOg4OL
...BLE NOTATION
- O M g 5: 3 VD 4eeO@ UL N #U h. Q.
ms Q> e i.
'O O A C Q. s. -
.atection (LLD) is defined in Table Notation a.
c""Lnd*
a.
The lower limit s.E'83"-
of Table 4.12-i 4 Specification 4.12.1.1.
e - - - C, C D e
C.OfM O.C # V
- a L.
1.,,,_.,
,,%,m.,
au vi--
1--
o f _1 t
C e-uo L
e m,,.
,m s :_
i--_3: A_
evn c 6.c 4 3 4 c a e vr 12---
-.jiw i uu a visu w i i vu a 11 1 w 7---
e m s-=
Vem
~.
r_. n s
2.
. uu i vi s u w 'i i u'n 1414 A LU I U b, 4 'L
- 4v v
'c p u a a 6 v 'a t -
v bV 6dcoaurT U R - U9 C) c=
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- anoai aaav uc ya i v i a,c u.viivnany aouovunn, 3Lorcup, u n
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u g C,- 8 f C.4 ruiu,y a.a gC
-;;;,;.;r opccasiccia. vpun ence whidi uvuiu signin cani.iy aitcc a
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B E O LaJ 4 V e
'O E I C
"O 4/5 6= e O E {,E g E{
d.
Cu. :n3.clueses vi; Pit r+ Hey.
- n s-C s.
Analyses shall also be performed at ic :t e.
ithi-13 'cu m W O. L e O.=e-L e e e O. C O
e.
4 following-e+eh shutdown, startup,orpil:r pe-etic-:!
ce n c.ca;;
- O r--
9 tt u p.u g.= g,
6 o m C s c>.C r.h cculd 10:d te ci n'ficant 4 m re:::: er de:rc;;;;
- r; die -
F 3., p 9
2 udia; cal; =::.
"':r ;;.r.plc: allcct,d f;r 2' '=r: 2r c uly:^ t K gy v)n m Ouso a 4
th: ;;..;;;;nding 'C ': ;j R i = c::e_ 27_; f =tcr cf !O:
- 3. F rf O
skalt lot, takw or least one, pex
{D kt;d). f.Trition hrab saaptasem. h re tit 4 tion akaust fr.~ %. spear fud.
j fp
~1 dup tot am ukemes spea pa. ts k L. gy g,14
,y The ratio of the sample flow rate to the sampled streem flow rate g
shall be known for the time period covered by each dose or dose fp',.
g.
rate calculation made in accordance with Specifications 3.11.2.1, p
5 3.11.2.2, and 3.11.2.3.
r h.
The principal gamma emitters for which the LLD specification will
?e apply are exclusively the following radionuclides:
Kr-87, Kr-88, g6 Xc-133, Je-133n, Xe-135, and Xe-138 for gaseous emissions and Mn-54, s.gl Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144
]f--
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 p g also be identified and reported. Nuul W=> nn sh o. c Lulem tN LLi 3
for th; naljscs shou!d et be repa,tcd : king prc=nt at tM tLE
's.:.1 for that c.ucl;de. W1.cr unute!
t=:
roult in LLO':
h e-1% mu!% e:== aim a e==tce - te 2 -
^ar:1 ef'lu;nt cport.-
LACBWR 3/4 11-9 4
l
\\
RAUl0ACl!VE LfflulflTS DOSE, fl0BLE GASES LIMITING CONDITION FOR OPERATION urrc;tricted ; m Occ rigure 5 ' l }- due to 3.11.2.2 The air dose i" noble gases released in gaseous effluents shall be limited to the following:
.from Ac. sHe, (su Figure. 5.1-l) a.
During any calendar qucett - to < 5 mrad for gamma radiation and < 10 mrad for beta radiationi b.
During any calendar year, to < 10 mrad for gama radiation and
< 20 mrad for beta radiation:~
-(The f o r -H ';" ^bj re t ite r.
-hall alan br r&wdbr'
- pccted publi; c: upancy cf arca;, c.g., ';cd-
" "i' ! !r-cc;ter: "i+hin the urer.tricted crc; bot,n d ry.
APPLICABILITY: At all times.
i^ hex o A9 oh ryorT u p; rect sph ii t: d i w b.4. h ACTION:
a.
With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits,4 prepare and submit to the Commission within 30_ days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to
. educe the releases of radioactive noble gases in gaseous effluents during the remainder of-the current calendar qgMdd dose duringtgg.and du subsequent three calendar quarters so that they.
these four calendar quarters is within (10) mrad for gamma radiation and (20) mrad for beta radiation.
b.
The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS cure 4T cabdv Suuter out ouvear caleada y 4.11.2.2.1 Dose Calculations Cumulative dose contrib&tions for th ti c Pri9 shall be determined in accordance with the Offsite Dose lcu-lation Manual (ODCM) at least once every 31 days.
- ,'1, p 9 ponnr+e Tho < miannien1 ondinac+4,e Erricen+ oeieece e p;, t-n 1
c h0 l ' 4 ".C l dc the.. f vi.q ti v. $pCCi#'cd
- ^ $' 2Cif4ts?" 6 9
'.9.
a j
LACBWR 3/4 11-10 r
DP c
- g-N j""
\\
a RADI0 ACTIVE EFFLUENTS DOSE, RADIOI0 DINES, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from radiciodines, radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days, in gaseous effluents released t-e -Jak sik rretric+ed 2rr- (see Figure 5.1-1) shall be limited to the following:
During any calendar quarter to < 7.5 mrem; a.
b.
linrinq any i.alendar year to - !!i mrem;
-(The dose des tyn objective shall be redut.ed based on pre licted T
carbon-14 releases if effluent sampling is not provided.)
I
^
f APPLICABILITY: At all times.
5 E
ACTION:
s T ~*
With the calculated dose from the release of radiciodines, a.
radioactive materials in particulate form, or radionuclides e
g.*7 ther than noble gases jn gaseous effluents exceeding any of e above limitQprepare and submit to the Commissicn within
~$
30 days, pursuant to Specification 6.9.2, a Special Report
/
which identifies the cause(s) for exceeding the limit and 3S defines the corrective actions to be taken to reduce the A
releases of radiciodines, radioactive materials in particu-half-lives greater than 8 day (other than noble gases) with late form, and radionuclides s, in gaseous ef fluents during the remainder of the current calendar quarto;rgtgrinqthe subsequent three calendar quarters so that tHeg._, r dese or dose.omitment to an individual from such releases during these f our calendar quarters is within (15) mrem to any organ.
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREIG TS cama catenaw qwter wL cwed cshudar s;su 4.11.2.3.1 Dose Calculations Cumulative dose contributions for the -te4L 3
tinc pcf..
siiall be determined in accordance with the ODCM at leact once every 31 days.
4.11.2.2.2 cre-te
'hc n-f annual C;dic;;tivc Efflucnt Cclene Deport c 'the t. e info..wi..en spcci' icd " Spc ification C.0.1.9.
S LACilWR
.1/1 11-11
-.a.su
.w _ -
=- ~. ~
. ~
RADI0 ACTIVE EFFLUENTS GASE0US RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION iot ofenttion.
The gaseous radwaste treatment syst2m shall be 0 E" FILE. The 3.11.2.4-?d:::tc tr^2tm,t ;yste ch;11 6 m3cd te, edm 4,radicccti cc g>eanne Swac"e e tc; prior to thei-dischcrg:.hc t, thc projccted
- ctericl: 4a 4-desc: dec to gcccca ef'!uent re!cc:c; ;c u.o c-guccu e f fi __n 2
stricted cccc; (;am Tiguce 5.1-1) whcc. a,ecugcd cscr 21 dcy:.;cu!" e.c M 0.2 rad fcc g;TTa redini.iun ec.d 0.4... cad for tctc. a d4 tfan APPLICABILITY:
St ' ' ' ' ' ' c;. Llhecew %. wta caede,s r air ejn t u spt ee.
it in opsention.
m.y,,au. %
16,. 1 <t,9, AC TIO!!:
W.416 6aitm RnDuAut ist.Amtat
.-m ee tw I;ea.} a q oik,< report r y* air. A q sp<id ;<,tinn b.ii. i,
rged fer nere the 21 ay+
d'
- a. g'i th gr,c; n.e t t ~ E ing
itNut trectm nt and i c::ces: cf th; ab7cc li ite, prepare and submit to the Comission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:
-tke kepemble er Identification of equipment 4 subsystems mot O'E"2'P 1.
A and the reason for nonoperability.
hoperalate, ncr cpc ab4 equipment to 2.
Action (s) taken to restore the3 OPERABLE STATUS.
3.
Sumary description of action (s) taken to prevent a recurrence.
h.
The provisions of Specification 3.0.3 and 3.0.4 are not applicable, 5URVEILLANCE RL_QUIREMINTS 4.11.2.4.1
%::: due
- n g uacus relec:c: to unrcst-icted aren cha!' be prcjected 2t 'eM+ ente p^- 21 dayt.
4.11.2.4.2 The gas radwaste treatment systems shall be demonstrated OPERABLEAat least once per 92 days unless the apprcprict:- system has been utilized to process radioactive gaseous effluents during the previous 92 days.
SV5 m 9twed _foy aX ut 6AsEcos RAD 4AsTC MAmW C->-
f cptsat;q lfe Tt _
rduTe.s)
! ACliWR 3/4 's-!?
I
{
N.*
DT"D P]O TyS' oJu M//L n
jj I$
RADI0 ACTIVE EFFLUENTS d
3 q mber.} h. %1h, b fo rah TOTAL
- s. V'
. DOSE 5$ J =4 f nuluctMy c4 vadin.t6, hron 3
~
LIMITING CONDITION FOR OPERATION i
5 r,
J e__?26 ba-j f* "3 [
The dose or dns} commitment to4 -rea' 5div%s4-4m-eH-3.11.2.o uranium f uel cycle source'.4 e limited to < 25 mrem to the total body or I.2 t F-6 i 3 ',I any organ (except the thyroid, which is ITmited to 1 75 mrem) over a f '.j y g ceriod of 12 consecutive months.
~.: e n
]'
APPLICABILITY: At all times.
5 +o ACTION _:
~
d c3 "
Withthecalculateddose{fromthereleaseofradioactive
$.f d a.
3 materials in liquid or gaseous effluents exceeding twice
[#Q=5tj the limi ts of Specifications 3.11.1.2.a, 3. ll. l.2.b, 3. ll.2.2.a,
3.11.2. 2.b, _3. ll.2.3.a, or 3. ll.2.3.b,Aprepare and submi t a a-t,
t y $.4 p Special Report to the4Ccris;i;n pursuan+ te S;cci'icatic-t 5
Vi C.S.: end li-it the sua;cqucat r.s.katc; cuch that +he dcce er c2 O j )
j d;;; c;Ht ent tn ' -: 1 indi. ; dual 'rc al' ura"f ur fuch
- f. f g
-1 eycle acccc; i
' N ted te 25 m m te c total body cr g{*
gj,j E ocgi.n (c; cept thyroid, dich o linnied te 75., c '
1 *.d 't
/ m cr 12 c;n;; cutie: =n ths This Speciai Report shall includa an analysis which demon:trctc; t':+gradiation exposures to a44 Yd
<$ f +o k3-gFre a l S d i " t : :Afrom a.14 uranium fuel cycle sources (including $oo g+ g t t
j all effluent pathways and direct radiation)Aare :cre +& thq F in u.tgg3
" O CI", P a c ' I M S t a d r.
0*" ~ ~ ire obt
- a " M :nt: 'rtm P.. -
d 53
- e. E d
12 0 CC- ! ;ich to pcrit relesse> nhhh mceds th: M CM J
d y L Part N O 'ada i b.
The prnvisions of Specification 3.0.3 and 3.0.4 are not J
applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5.1 Dose Calculations Cumulative dose contributions from liquid and caseous effluents shall be determined in accordance with Specifications 4 4
- 2. l f i. 2. e, 3.11.1. 2. b. 3. ". ?. 2. ;, 3.11. 2. 2. 5, 3.11. 2. 3. 0, Md 3 ".2.3 A and in accordance with the Offsite Dose Calculation Manual (00"'i).
N 4.ll.2.5.2
[yOrt0 Sp9C CECit3 3hOll UC 3ub>Mtted 03 FCNI 3 U E bI' b
I'I U
-9pe rft atiin 2 M c A
+
& f or a 12 rnnsecutive month peri.nl t hat i:-
k.~
includes t he role.e.o(.) covered by this report.
If the estimated i.
dnse(s) exceeds the limits of Specific.ition 3.11.4, and if the release L
condition resulting in violation f 40 Cl R 190 has not already been corrected, the Special Report shall include a request for a variance Y
in accordance with the provisions of 40 CFR 190 and including the specified information of 6 190.11(b).
Submittal of the report is
?
considered a timely request, and a variance is granted until staff P
- action on the request is complete. The' variance only relates to the W
limits of 40 CFR 190, and does not apply in any way to the requirements J-for doce limitation of 10 CFR Part 20, as addressed in other sections of this technical specification.
RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING COND'ITION FOR OPERATION 3.11.2.6 The concentration of hydrog in the main condenser offgas treatment system shall be limited to <y4% by volume. fter recombination, 2.
APPLICABILITY: At all times.
ACTION:
A[}(Pi+k coacent-ation cf hydrcgen 4-the
~'4" cenden;;r of# gat
+ha trect:::nt syster exceeding th; limit, restcr; th:
- ncentraticn to "4 thia the 'i-it "ith 13 Scurc_
c h.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.6 The concentration of hydrogen in the main condenser offgas treatment system shall be determined to be within the above limits by continuously monitoring the waste gases in the main condenser offgas treatment system with the hydrogen monitors required OPERABLE by Table 3.3-12 of Specification 3.3.3.9.
With the concentration of hydrogen 'ad/^r crig:n in the main condenser of fgas treatment system greater than 2% by volume but less than or equal to 4% by volume, restore the concentration of hydrogen end/cr eny;e-to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
b.
With the concentration of hydrogen 2nd/cr c:a gri in the main condenser offgas treatment system greater than 4% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of hydrogen :n?/sr ::g;c-to less than or equal to 2%
within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
LACBWR 3/4 11-14
I b
RADI0 ACTIVE EFFLUENTS 1
g
,3/4.11.3 SOLID RADIO.TCTIVE WASTE
+, a applicate. is ocurAAuc. 44 o P90cr.ss courtou nosw, k
LIMITING CONDITION FOR OPERATION l
3 The solid radwaste system shall be OPERABLE and us)A c prev %
t 3.11.3.1
~for theApackaging of & radioactive wastes 4 to ensure ek meeting -e4 the require.aents of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment g
of radioactive wastes from the site.
3 h,j APPLICABILITY: At all times.
ACTION:
l' o
W%
art 20 and 10 CFR Part 71 not
[T a.
With the3 requirements of 10 CFR 3
satisfied, suspenc; shipments of
^#^'+i ^ ^ ~ " + ' * "~ " s ol i d T.T radioactive wastes from the site.
de Miv@ ra Q A
- inopeaMa, f f, b.
With the solid radwaste systemget 0"9'"'l for more than 31 1.:
,g b days,Jgder "equired te =ct 7 0.FR--Isrt 20 and 10 CIP, Pe. u ih prepare and submit to the Commission within 30 days, pursuant Q
to Specification 6.9.2, a Special Report which includes the following information:
-Se A eper ato or Identification ofa quipment e4 subsystems net OPEn' LL and 1.
e 3
the reasons for inoperability.
2.
Action (s) taken to restore the inoperable equipment to OPERABLE status.
4ke souosicATud a rad ion.tt<e, 3.
A description of alternative used for packaging o wastes.
3 3
4.
Sunmary description of action (s) taken to prevent a recurrence.
The provisions of Specifications 3.0.3 and 3.0.4 are not c.
applicaole.
SURVEILLANCE REQUIREMENTS The salid ga,dwaste system shall be demonstrated OPERABLE at 4.11.3.1.1
'cr ther: be th: ::p bi'ity #c pech ging ;f wa;t least once per 92 days 3 by ;ccting ei.e or mere c' t% condi tiens bci;..-
Oj pcrfcm:n;c of f=cticn;1 tc;t: cf the cg.ip cnt nd ;;;poncata r
m.--
_u;
_,.&.,,..m.
e_ j... _
vs v a a w.
.a v a su AE. 4y operating the solid radwaste system at least once in the previous 92 days /is accorda co a:A A PgocEss coonot pao 6tw, er b.e' Verification of the existence of a valid contract for p:Wgin;.
SoupipiceTina to be. per{wuA lof a. tedractor 6 accorduw wi4k l
/
Q,5 LAC 8WR gcg gg p 9
l SURVEILLANCC REQUIREMENTS - (Cont'd) c;;rt:
TS: :: ::r,ucl ";di;;;tive Effluent a:!::se rscgv.t 4.11.3.1.3)p a
Sh0ll include one TollDWing infei satiOr fOr 000h ty, c vf 55 lid na0tc
_rbipp:d effa.tc daria; the.cr.-t pacicdr e
,1.
Cont;iGCr JOIL:C, tet:1 Curic qu;ntit, (detc.aincd by ;;;;;r;m:nt Or 0;;io.uic).
b.
C.
pr#r. ip;l g;r 2 *edianue14 der (deter :n;d b, meuau,u,,,unt 0,-
r-
,+,,
ms..
~._,,
f type c< e m ce_:.. :p =t raia, = c n : J.,
nu a,
e.
cvaporate-betten:),and-a.
type nf enn+'ine- (e_
L(a Type 8 Tyra R, ' '":2 Qu?"t'ty).
THE PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICA-L 5
TION'cf at least one representative test specimen from at least every tenth batch of each type of wet radioactive was~e fe.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).
If any test specimen fails to verify SOLIDIFICATION, the <0LIDIFICA-TION of the batch under test shall be suspended until such time as a.
additional test specimens can be obtained, alternative SOLIDIFICATION parametars can be determined in accordance with the PROCESS CONTROL SOLIDIFICATION PROGRAM, and a subsequent test verifies SOLIDIFICATION.
of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.
If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the b.
collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least The 3 consecutive initial test specimens demonstrate 50LIO1FIC TION.
PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste.
3/4 11-16 LACBWR
i 5
INSTRUMENTATION BASES C}
3/4.3.3.8 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION
]
The radioactive liquid effluent instrumentation is provided to monitor e
and control, as applicable, the releases of radioactive materials in a-f!
liquid effluents during actual or potential releases of liquid effluents.
4 The alarm / trip _setpoints for these instruments shall be calculated in accordance withAM"C Oppre"ed :t:.ede in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
f 3/4.3.3.9 RADI0 ACTIVE GASEOUS "pOC:;S AND-EFFLUENT MONITORING 4
INSTRUMENTATION 2
The radioactive gaseous effluent instrumentation is provided to monitor l-and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.
31 The alarm / trip setooints for these instruments shall be calculated in A
accordancewitMN"I:ppr=d;c3sd;intheODCMtoensurethatthe
"~
alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.
The p e-^: -monitoring instrumentation includes provisions for monitoringj[~ly the concentrat'Ons of potentially explosive gas mixtures in the main condenser offgas treatment system.
The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
e y,
o E
LACBWR 8 3/4 3-4
~
_ :=
' L L a w..
.. x :q
- g. _x
-.m-- m
3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 5
.5 3/4.11.1.1 CONCENTRATION
,4. h 2.,
This specification is provided to ensure that the concentration of radio-
[
active materials released in liquio waste effluents from the site 4e-o
= restricted ;r:2: will be less than the concentration levels specified 3
in 10 CFR Part 20, Appendix B, Table IIA This limitation provides addi-2 tional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures within (1) the 4
Section II. A design objectives of Appendix I,10 CFR Part 50, to an 4
indivMual and (2) the limits of 10 CFR Part 20.106(e) to the population.
The cocentration limit forinoble gases is based upon the assumption that i
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 Comission on Radiological Protection (ICRP)
Publication 2.
3/4.11.1.2 DOSE 4
4p This specification is provided to implement the requirements of Sections c.
II.A III. A and IV.A of Appendix I,10 CFR Part 50.
The Limiting Condition for Operation implements the guides set forth in Section II.A f
g,Ps of Appendix I.
The ACTION statements provide the required operating flexib_ility and at the same time implement the guides set forth in d,
Secti~onkH-A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably y
achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reason-
<j g 32 able assurance that the operation of the facility will not result in wj radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calcuhl,tions in the s
ODCM implement the requirements in Section III.A of Appendix I that d 'r confomance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an
-g-i individual through appropriate pathways is unlikely to be substantially
., J underestimated.
The equations specified in the ODCM for calculating jj the doses due g the actual release rates of radioactive materials in be consistent with the methodology provided in liquid effluents H
Regulatory Guide ^W.109, " Calculation of Annual Doses to Man from Routin y
o* Y 1
}with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113 " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I,"
April 1977. NUE G-G133 g,u<ida e heds for dere ca hu h tier.: : r.;i tcnt vith ",c @le;;g Cuidc; 1.100 nd 1.112.
LACBWR B 3/4 11-1
O 8
RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1.3 LIQUID WASTE TREATMENT Tne OPERABILITY of the liquid radwaste treatment system ensures that this sg system will be available for use whenever liquid effluents require treat-w'{
ment prior to release to the environment.
The requirements that the appropriate portions of this system be used when specified provides assur-g ance that the releases of radioactive materials in liquid effluents will
.P j@
be kept "as low as is renonably achievable." This specification imple-ments the requirements of 16 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section 11.0 of
-d Appendix I to 10 CFR.Part 50. The specified limits governing the use of 4
appropriate portions of the liquid radwaste treatment system were speci-fiedasasuitablefractionoftheyw44e.setforthinSectionII.Aof Appendix I,10 CFR Part 50, for liquid effluents.
3/4.11.2 GASE0US EFFLUENTS 3/4.11.2.1 DOSE RATE 4-Thgspecification is provided to ensure that the dose rate at anydime at theg = E;ico crc boundary from gaseous effluents from all Units on the c
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 concentra-tions of 10 CFF Part 20, Appendix B, Table II' These limits provide 3
reasonable assurance' that radioactive material discharged in gaseous P
effluents will not result in the exposure ogan individual in an unre-f r'ric-crc boundary, stricted area, either within or outside thee to annual average concentrations exceeding the limits specified in g
Appendix B, Table II' of 10 CFR Part 20 (104FR Part 20.106(b)).
For individuals who may at times be within thdy='"e4m re: boundary, the occupancy of the. individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the e site, k ;i c r.
~2 boundary.
The specified release rate limits restrict, at all times, the corresponding ga nd beta dose rates above background to an i ividual at or beyond th 5%eir crc: boundary to < 500 mrem /
year t t total body or to < 300 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
< 1500 mrem / year for the nearest cow to the plant.
1 LACBWR 8 3/4 11-2 u-.
m J
RADI0 ACTIVE EFFLUENTS BASES
[
',11. 2. 2 DOSE, N0BLE GASES This specification is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10 CFR Part 50.
The Limiting Condi-tion for Operation implements the guides set forth in Section II.8 of Appendix I.
The ACTION statements provide the required operating flexi-bility and at the same time implement the guides set forth in Section IV.A of Appendix I$ assure that the releases of radioactive material in gaseous The Surveil-effluents will be kept "as low as is reasonably achievable."
lance Requirements implement the requirements in Section III.A of Appendix 1 that conformance with the guides of Appendix I is to be shown by calcu-lational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be sub-stantially underestimated. The dose calculations established in the ODCM for calculating the doses due tLthe actual release rates of radioactive noble gases in gaseous effluentsp? 5: consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-CooledReactors," Revision 1,Julyig7. TheODCMequationsprovidg
% ased for determining the air doses at the chrien ame boundary wi!! S b
upon the historical average atmosphekc conditions.
'i""EC C! ?2 p re c i &^
00 and
=th:2 Sr 2:0 akuktion; conci; tent ith " phte y Guit:
h+M.
3/4.11.2.3 DOSE, RADIOI0 DINES, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES This specification is provided to i.nplement the requirements of Sections II.C, III.A, and IV.A of Appendix I,10 CFR Part 50.
The Limiting Conditions for Operation are the guides set forth in Section II.C 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 A;pendix I to assure that the releases of radioactive materia s in gaseous ef fluents will be kept "as low as is reasonably achieva The ODCM calculational methods specified in the surveil-lance 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 mod;1s and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.
The ODCM calculational methods app
/cd by NRf-for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision I, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and l LACBWR 8 3/4 11-3 mW44.p%*. W %d'MJ* Q fA..'.,.4 i % 1 ## J 4.
J M4M
RADI0 ACTIVE EFFLUENTS BASES Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.
These equations also provide for detemining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radiciodines, radioactive material in particulate fom and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area.
The pathways which are examined in the develop-ment of these calculations are:
(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 mean producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man.
3/4.11.2.4 GASEOUS WASTE TREATMENT _
y The OPERABILITY of the gaseous radwaste treatment system ensures that the system will be available for use whenever gaseous effluents require treat-
-f
'ent prior to release to the environment.
The requirement that the appropriate portions of the system be used when specified provides reason-y 9
able assurance that the releases of radicactive materials in gaseous j
effluents will be kept "as low as is reasonably achievable." This speci-fication implements the requirements of 10 CFR Part 50.36a, General Design a
Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section IID of Appendix I to 10 CFR Part 50. The specified limits governing the C
use of appropriate portions of the systems were specified as a suitable f raction of thege+i4e set forth in Sections II.B and 11.C of Appendix I, 10 CFR Pert 50, f or gaseous effluents.
TOTE 3/4.11.2.5, DOSE
" " " " " #" " W
- Pac 4 i _
3/4.11.2.6 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flaninability limits of hydrogen and oxygen. Automatic control features are included in the system to pre-vent the hydrogen concentrations from reaching these flammability limits.
These automatic control features include containment of the hydrogen anL automatic diversion to recombiners.
Maintaining the concentration of hydrogen below the flammability limits provides assurance that the releases of radioactive materials will be controlled in confomance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
LACBWR B 3/4 11-4
0 e
RADIOACTIVE EFFLUENTS BASES 3/4.11.3 SOLID RADI0 ACTIVE WASTE The OPERABILITY 0* the solid radwaste system ensures that the system will be available for use w*.anever solid radwastes require processing and packaging prior to being shipped offsite.
This specification im lements the requirements of 10 CFR Part 50.36a and General Design Crit >
60 of Appendix A to 10 CFR Part 50. h process puame.iers Ac6ded. L utal,tisL'q
.< e. ut t;=iteA % waste. iye, h fToce.ss Codilta pgessew q AcL.L, ( ct Maste. pH, Maste} gat / so(MehcA;o~ qeJ/ catalyst ratios, Mcste oil hic 4 faaste. princtpte. ekeutcal cwsttiac~ts, mixi$ a d hviq tlw.s,.
3/4.11. 5 TOTAL DOSE
~
This specification is provided to meet the dose limitations of 40 CFR 190.
The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive ef tluents exceed twice the design objective doses of Appendix I.
For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a nember of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level.
The Special Report will describe a course of action which should result in the limitation of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits.
For the purposes of the Special Report, it may be assumed that the dose commit-ment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be con-sidered.
If the dose o any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staf f action is completed.
An individual is not considered a member of' the public during any period in which he/she is engaged in carring out any operation which is part of the nuclear fuel cycle.
, LACBWR B 3/4 11-5 l
)
t.
~*
f i
_3/4.0 AP__P_LI CAB _._I_LITY l
. LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the limiting Conditions for Operation contained-in the succeeding Specifications is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
j 3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals.
If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, the unit shall be placed in a CONDITION in which the Specification does not apply by placing it, as applicable, in:
r 1.
At least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 2.
At least COLO 5HUIDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Wnere corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time l
limits as measured from the time of failure to meet the Limiting Condition for Operation.
Exceptions to these requirements are stated in the individual Specifications.
3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made unless the conditions for the Limiting Condition fer C'peration are ret without reliance on provisions contained in tne ACTION receircments.
This r
i prevision shall not prevent passage thrcugh OPEP.ATIONAL C00ili0NS as reovired l
to comply witn ACTION recuirements.
Exceptions to these receirements are stated j
in the individual Specifications.
4 5.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency po,eer source is inoperable, or sciely because its normal < power source is inoperable, it may be considered C'EIAE' E fcr the purpose of satisfying the receirements of its ap;1icable limiting condition for Operation provided:
(1) its correspondine normal or ec+rgency power source is '0PERABLE; and (2) all of its redundant system (s),
oystem(s), train (s), component (s) and device (s) are OPERASLE, or likeviise scsatisfy the requirements of this specification.
Unless both conditions (1)
.and (2) are satisfied, the unit shall be placed in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COLD SHUT 00WN'within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
~
This specification is not applicable in OPERATIONAL CONDITION 4 or 5.
.3/40-3
- 6. 0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The (Plant Superintendent) shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.
6.1. 2 The Shift Supervisor shall be responsible for the Control Room command function, and shall be the only individual that may direct the licensed activi-ties of licensed operators.
A management directive to this effect, signed by the (highest level of corporate mcnagement) shall be reissued to all station personnel on an annual basis.
6.2 ORGANIZATION 0FFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown on Figure 6.2.1-1.
UNIi $TacF 6.2.2 The Unit organization shall be as shown on figure 6.2.2-1 and:
a.
Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.2-1.
b.
At least one licensed Reactor Operator other than the Shift Supervisor shall be in the control room when fuel is in the reactor.
c.
At least two licensed Reactor Operators other than the Shift Supervisor shall be present in the control room during reactor start up, scheduled reactor shutdown and during recovery from reactor trips.
d.
A health physics technician
- shall be onsite when fuel is in the reactor.
e.
All CORE ALTERATIONS shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no cther concurrent responsibilities during this operation.
f.
A site Fire Brigade of at least 5 members shell be maintained onsite at all times *.
The Fire Brigade shall not include (3) members ~of the minimum shift crew necessary for safe shutdown of the unit and any personnel -required for other essential functions during a fire emergency.
wThe health physics technician and Fire Brigade compositior may be less than l
the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.
GE-STS 6-1
-w,.
2 u-m.
4 4
+
i i
i j
i j
This figure shall show the organizational structure and lines of responsibility for the offsite groups 2
that provide technical and management support for the unit..The organizational arrangement for performance and monitoring Quality Assurance activ-ities should also be indicated.
i l.
i-Figure'6.2.1-1 i
0FFSITE ORGANIZATION GE-STS-6-2 L
v
,.1_A.---.
,e
j l
l i
This figure shall show the organizational structure and lines of responsibility for the unit staff.
Positions to be staffed by licensed personnel should be indicated.
i Figure 5.2.2-1 UNIT ORGANIZATION GE-STS 6-3
ABLE 6.2.2-1 MINIMUM SHIFT CREW COMPOSITION SINGLE UNIT FACILITY (Optional) 1 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2, & 3 CONDITIONS 4 & S SS 1
1
.SRO 1
None R0 2
1 A0 2
1 STA 1
None TWO UNITS WITH TWO SEPARATE CONTROL ROOMS (Optional)
WITH' UNIT (2) IN CONblTION 1, 2, OR 3 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 a
a SS l
y SRO 1
None R0 2
1 A0 2
1 a
l STA l
None WITH UNIT (2) IN CONDITION 4 OR 5 OR DE-FUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS'1, 2, & 3 CONDITIONS 4 & 5 a
a SS l
y SRO 1
None R0 2
l b A0 2
2 STA 1
None GE-STS 6-4a i
o TABLE 6.2.2-1 (Continued)
MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH'A COMMON CONTROL ROOM (Optional) e I
WITH UNIT (2) IN CONDITION 1, 2, OR 3 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 a
a
.55 l
y a
SR0 l
None b
R0 2
1 b
A0 2
1 a
STA l
None
~
WITH UNIT (2) IN CONDITION 4 OR 5 OR DE-FUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 a
a SS l
y SR0 1
None R0 2
1 b A0 2
2 STA 1
None I
a GE-STS
'6-4b
'v._
3 i
i
TABLE 6.2.2-1 (Continued)
MINIMUM SHIFT CREW COMPOSITION TABLE NOTATION a/ ndividual may fill the same positior on Unit 2.
I N ne of the two required individuals may fill the same position on Unit 2.
0 SS - Shift Supervisor with a Senior Reactor Operators License on Unit (1).
SRO - Individual with a Senior Reactor Operators License on Unit (1).
R0 - Individual with a Reactor Operators License on Unit (1).
AO - Auxiliary Operator STA - Shift Technical Advisor Except for the Shift Supervisor, the Shift Crew Composition may be one less than the minimum requirements of Table 6.2.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2.2-1.
This provision does not permit any shift crew position to be unmanned upon shift change due to an onr^ ming shift crewman being late or absent.
The Shift Supervisor shall maintain his normal work station in the Control Room.
During any absence of the Shift Supervisor from the Control Room, an individual
~
(other than the Shift Technical Advisor) with a-valid SRO license shall be designated as the Shift Supervisor and shall assume the Control Room command function.
Licensed operators shall*:
1.
Not work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> straight, 2.
'Have at least a 12-hour break between work periods, 3.
Not work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in'any 7-day period, and 4.
Not work more than 14 consecutive days without having 2 consecutive days _off.
- Deviation from these requirements may be authorized by the (Plant Superintendent) in accordance with established procedures and with documentation of the cause.
4 s
GE-STS 6-4c
f ADMINISTRATIVE CONTROLS 6.2.3 (NUCLEAR EXPERIENCE REVIEW PANEL) i 6.2.3.1 The (Nuclear Experience Review Panel) (NERP) shall be a multidiscipline review group and shall review nuclear industry operational experience.
' 6.2.3.2 The (NERP) shall inform the-Shift Technical Advisor of all such experience that is relevant to operation of the unit.
6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall serve in an advisory capacity to the Shift Supervisor on matters pertaining to the engineering aspects assuring safe operation of the unit.
6.2.4.2 Ths shift Technical Advisor shall disseminate relevant operational experience identified by the (NERP).
. 6.3 UNIT STAFF QUALIFICATIONS Ilinimumqualificationsformembersoftheunitstaffmaybespecifiedbyuseof an overall qualification statement referencing ANSI N18.1-1971 or alternately by specifying individual position qualifications.
Generally, the first method is preferable; however, the second method is adaptable to those unit staffs requiring special cualification statements because of a unioue oroanizational structure.
6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI N18.1-1971 for comparable positions, except for (1) the (Radiation l
Protection Manager) who shall meet or exceed the qualifications of Regulatory
- Guide 1.8, September 1975 and (2) the Shif t Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with transientsandaccidents.gesign,andresponseandanalysisoftheplantfor specific training in plan 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the (position title), shall meet or exceed the l
requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55, shall include familiarization with relevant industry operational experience identified by the (NERP), and shall include degraded core training.
6.5 REVIEW AND AUDIT The method by which independent review and audit of facility operations-is accomplished may take one of several forms.
The licensee may either assign this functiva to an organizational unit separate and independent from the group having responsit,ility for unit operation or may utilize a standing committee composed of-individuals from within and outside the licensee's organization.
Irrespective of the method used, the licensee shall specify the details of each functional element provided for the independent review and audit process as illustrated in the followino example sJLe_c_ifications.
pas an interim measure until January, 1981, the Shift Technical Advirer function may be' performed by an SRO who augments the shift manning.r SR0's.)
GE-5TS 6-5 n
o ADMINISTRATIVE CONTROLS 6.5.1 UNIT REVIEW GROUP-(URG)
FUNCTION 6.5.1.1 The (Unit Review Group) shall function to advise the (Plant Superintendent) on all matters related to nuclear safety.
COMPOSITION
- 6. 5.1. 2 The (Unit Review Group) shall be composed of the:
Chairman:
(Plant. Superintendent)
Member:
(Operations Supervisor)
Member:
(Technical Supervisor)
Member:
(Maintenance Supervisor)
Member:
(Plant Instrument and Control Engineer)
Member:
(Plant Nuclear Engineer)
Member:
(Health Physicist)
ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the (ORG)
Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in (URG) activities at any one time.
MEETING FREQUENCY 6.5.1.4 The (URG) shall meet at least once per calendar month and as convened by the (URG) Chairman or his t'esignated alternate.
QUORUM 6.5.1.5 The minimum quorum of the (URG) necessary for the performance of the (URG) responsibility and authority provisions'of these Technical Specifications shall consist of the Chairman or his designated alternate and four members 1ccluding alternates.
RESPONSIBILITIES
- 6. 5.1. 6 The (Unit Review Group) shall be responsible for:
a.
Review of (1) all procedures required by Specification 6.8 and changes thereto, and (2) any other proposed procedures or changes thereto as determined by the (Plant Superinter. dent) to affect nuclear safety, b.
Review of all proposed tests and experiments that affect nuclear safety, c.
Review of all proposed changes to Appendix "A" Technical Specifications.
d.
Review of all pr(josed changes or modifications to unit systems or equipment that affect nuclear safety.
GE-STS 6-6
c ADMINISTRATIVE CONTROLS
~ RESPONSIBILITIES (Continued)
Investigation of all violations of the Technical Specifications e.
including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the (Superintendent of Power Plants) and to the (Company Nuclear Review and Audit Group).
f.
Review of events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written notification to the Commission.
g.
Review of unit operations to detect potential nuclear safety hazards.
h.
Performance of special reviews, investigations or analyses and reports thereon as requested by the (Plant Superintendent) or the (Company Nuclear Review and Audit Group).
i.
Review of the Security Plan and implementing procedures and shall submit recommended changes to the (Company Nuclear Review and Audit Group).
j.
Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the (Company Nuclear Review and Audit Group).
AUTHORITY I
- 6. 5.1. 7. The (Unit Review Group) shall:
Recommend in writing to the (Plant Superintendent) approval or a.
disapproval of items considered under 6.5.1.6(a) through (d) above.
b.
Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an j
i unreviewed safety question.
c.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the (Superintendent i
of Power Plants) and the (Company Nuclear Review and Audit Group) of disagreement between the (URG) and the (Plant. Superintendent); however, the (Plant Superintendent) shall have responsibility.for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 4
- 6. 5.1. 8 The (Unit Review Group) shall maintain written minutes of-each (IRG) meeting that, at a minimum, document the results of all (URG) activities performed under the responsibility and authority provisions of these Technical Specifications.
Copies shall be-provided to the (Superintendent of Power Plants) and the (Company.
Nuclear Review and Audit Group).
GE-STS 6-7
' ADMINISTRATIVE CONTROLS 6.5.2 COMPANY NUCLEAR REVIEW AND AUDIT GROUP (CNRAG)
FUNCTION 6.5.2.1 The (Company Nuclear Review and Audit Group) shall function to provide independent review and audit of designated activities in the areas of:
a.
nuclear power plant operations b.
nuclear' engineering c.
chemistry and radiochemistry d.
metallurgy e.
instrumentation and control f.
radiological safety g.
mechanical and electrical engineering h.
quality assurance practices i.
(other appropriate fields associated with the unique characteristics of the nuclear power plant)
COMPOSITION 6.5.2.2 The'(CNRAG) shall be composed of the:
Director:
(Position Title)
Member:
(Position Title)
Member:
(Position Title)
Member:
(Position Title)
Member:
fPosition Title)
ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the (CNRAG)
Director to serve on a. temporary basis; however, no more than two alternates shall participate as-voting members in (CNRAG) activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the (CNRAG) Director to provide expert advice to the (CNRAG).
GE-STS-6-8 I
_~
ADMINISTRATIVE CONTROLS b
MEETING FREQUENCY
- 6.5.2.5 The (CNRAG) shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per six months thereafter.
QUORUM
- 6.5.2.6 The minimum quorum of the (CNRAG) necessary -for the performance of the (CNRAG) review and audit functions of these Technical Specifications shall consist of the Director or his designated alternate and (at least 4 CNRAG) members including alternates.
Na more than a minority of the quorum shall have line responsibility for operation of the unit.
REVIEW-6.5.2.7 The (CNRAG) shall review:
a.
The safety evaluations for 1) changes to procedures, equipment or.
systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
e b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
c.
Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59,'10 CFR.
d.
Proposed changes to Appendix A Technical Specifications or this Operating License, e.
Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety.
g.
Events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written notification to the Commission.
h.
All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety.
i.
Reports and meetings minutes of the (Unit Review Group).
GE-STS 6-9
ADMINISTRATIVE CONTROLS-AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the (CNRAG).
These-audits shall encompass:
a.
The conformance of unit operation to provisions contained within the Appendix A Technical Specifications and applicable license conditions at least once per 12 months.
b.
The performance, training and qualifications of the entire unit staff at least once per 12 months.
c.
The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.
d.
The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.
e.
The Emergency Plan and implementing procedures at least once per 24 months.
f.
The Security Plan and implementing procedures at least once per 24 months.
g.
Any other area of unit operation considered appropriate by the (CNRAG) or the (Vice President Operations).
h.
The Fire Protection Program and implementing procedures at least once per 24 months.
i.
An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either-qualified offsite licensee personnel or an outside fire protection firm.
j.
An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 3 years.
AUTHORITY 6.5.2.9 The (CNRAG) shall report to and advise the (Vice President Operations) on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.
GE-STS 6-10
ADMINISTRATIVE CONTROLS RECORDS 6.5.2.10 -Records of (CNRAG) activities shall be prepared, approved and distributed as indicated below:
a.
Minutes of each-(CNRAG) meeting shall be prepared, approved and forwarded to the (Vice President-Operations) within 14 days following each meeting.
b.
Reports of revietss encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the (Vice President-Operations) with'n 14 days fol'.? wing completion of the review.
Auait reports encompassed by Section 6.5.2.8 above, shall be forwarded c.
to the_(Vice President-Operations) and to the management positions responsible for the areas audited withir 30 days after completion of the audit.
s l
6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:
(
a.
The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.
b.
Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission shal'. be reviewed by the (URG) and submitted to the (CNRAG) and the (Superintendent of Power Plants).
- 6. 7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
a.
The unit shall be placed it, at least HOT SHUTDOWN within two hours.
b.
The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour.
The (Superintendent of Power Plants) and the (CNRAG)'shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
A Safety Limit Violation Report shall be prepared.
The report shall be reviewed by the (URG).
This report shall describe (1) applicable circumstances-preceding the violation, (2) effects of the violation-upon unit components, systems or structures, and (3) ' corrective action taken to prevent: recurrence.
d.
The Safety Limit l Violation Report shall be submitted to the Commission, the (CNRAG) and the (Superintendent of Power Plants) within 14 days i
of the violation.
l GE-STS 6-11 e
c,
ADMINISTRATIVE CONTROLS 6.8-F30CEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
The applicable procedures recommended in Appendix "A" of Regulatory a.
Guide 1.33, Revision 2, February 1978.
4 b.
Refueling operations.
c.'
Surveillance and test activities of safety related equipment.
d.
Security Plan implementation.
e.
Emergency Plan implementation.
f.
Fire Protection Program implementation.
6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the-(URG) and approved by the (Plant Superintendent) prior to implementation and re ieweii periodically as set forth in administrative procedures.
6.8.3 Tsoporary changes to. procedures of 6.8.1 above may be made provided:
a.
The intent of the original procedure is not altered.
b.
The change is approved by two members of the unit management staff, at least one of whom holds a Senior Reactor Operator's License'on the unit affected.
c.
The change is documented, reviewed by the (URG) and approved by the (Plant Superintendent) within 14 days of implementation.
6.9 REPORTING REQUIREMENTS
_ ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement, unless otherwise noted.
STARTUP REPORT 6.9.1.1 A_ summary report of plant startup and power escalation testing shall be submitted following (1): receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel suppiier, and (4) modifications that may have significantly altered the nuclear, 1
thermal, or hydraulic performance of the unit.
-GE-STS 6-12 t
- A'DMINISTRATIVE CONTROLS
- STARTUP REPORT (Continued)
.6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall.. include a description of the measured values of the operating conditions cr characteristics obtained during the test program and a comparison of these values with design predictions'and specifications.
Any corrective actions that were required to obtain satisfactory operation shall also be
~
described.
Any additional specific details required in license conditions based on othe commitments shall'be~ included in this report.
6.9.1.3 Startup reports shall be submitted within (1) 90 days following.comple-tion of the startup test program, (2) 90, days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.
If the Startup Report does not cover all three events, i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation, supplementary reports shall be submitted at least every three months until all three events have been completed.
M ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous cdendar year shall be submitted prior to March 1 of each year.
The initial report shall be submitted prior to March 1 of the year following initial criticality.
6.9.1.5 Reportsrequiredobanannualbasisshallinclude:
a.
A tabulation on an annual basis of the number of station, utility, and other personnel, including contractors, receiving exposures greater work ~andjobfunctions,gressociatedmanremexposureaccordingtoe.g.,
than 100 mrem /yr and th inservice inspection, routine. maintenance, special maintenance (describe maintenance), waste processing, and refueling.
The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements..Small exposures totalling less than 20 per-1-
cent-of the individual total dose need not be accounted for.
In the aggregate, at least 80 percent of the total whole body dose received from external scarces should be assigned to specific major work functions.
b.
(Any other unit unique reports required on an annual basis).
M3NTHLY OPERATING kEPORT 6.9.1.6 Routine reports of. operating statistics and shutdown experience, includ-ing documentation of all challenges to (safety valves o'r) safety / relief valves,
~
i 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 copy to-the Regional Office of Inspection and Enforcement, no later than the 15th of.each month following the calendar month covered by the report.
~
~
17A single submittal may be~ made for a multiple unit station.
The submittal should combine'tiose sections that are common'to all units at the station.
EThis tabulation supplements the requirements of 620.407 of_ 10 CFR Part 20.
GE-STS-6-13 4
1 t
t
~ADMINI'STRATIVE CONTROLS REPORTABLE OCCURRENCES I
6.9.1.7 The REPORTABLE OCCURRENCES of Specifications 6.9.1.8 and 6.9.1.9 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. ' Supplemental reports may be required to fully describe final resolu-tion of occurrence.
In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.
PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.8 The types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designate no later than the first working day following the event, with a written followup report within 14 days.
The written followup report shall include, as a minimum, a completed copy of a licensee event report form.
Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the ci.cumstances surrounding the event.
a.
Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a moaitored parameter reaches the setpoint speci-fied as the limiting safety system setting in the technical specifica-tions~or failure to complete the required protective function.
b.
Operation of the unit cr affected systems when any parameter or opera + ion subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation _ established in the technical specifications.
c.
Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.
d.
Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power opera-tion greater than or equal to 1% delta k/k; a calculated reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an' unplanned reactivity insertion of more than 0.5% delta k/k; or occurrence of any unplanned criticality.
~
e.
Failure or' malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional require-ments of system (s) used to cope with accidents analyzed in the SAR.
f.
Personnel error or proc'edural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.
GE-STS 6-14
.ADMINISTRA,TIVE_ CONTROLS PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP (Continued) g.
Conditions arising from natural or man-made events that, as a direct result of *.le event, require unit shutdown, operation of safety systems, r,. Other protective measures required by technical specifications.
h.
Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.
i.
- Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner-less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial action or corrective measures to prevent the existence or development-of an unsafe condition.
THIRTY DAY WRITTEN REPORTS
- 6. 9.1. 9 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occurrence of the event.
The written report shall include, as a minimum, a completed copy of a licensee event report form.
Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
a.
Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.
b.
Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.
c.
Observed inadequacies in the implementation of administrative or proce-dural controls which threaten to cause reduction of degree of redundancy
-provided in reactor protection systems or engineered safety feature systems.
d.
Abnormal degradation of systems other tlan those specified in
-6.9.1.8.c above designed to contain radSactive material resulting from the fission' process.
t i
GE-STS' 6-15 i
i
ADMINISTRATIVE CONTROLS SPECIAL REPORTS..
~
Special reports may be required covering inspections, test and maintenance activ-ities.. These special. reports are determined on an individual basis for each unit and their preparation and submittal are designated in the. Technical Specifi-cations.
6.9.2 Special reports sball be' submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each' report.
6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations,' the following records shall be retained for at least the minimum period indicated.
6.10.1 The following records shall be retained for at least five years:
a.
Records and logs of unit operation ccvering time interval at each power level.
b.
Records and: logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c.
ALL REPORTABLE OCCURRENCES submitted to the Commission.
d.
Records of surveillance activities, inspections and calibrations required by these: Technical Specifications.
e.
Records of changes made to the procedures required by Specification
-6.8.1.
f.
Records of radioactive shipments.
g.
Records of sealed' source and fission detector leak tests and results.
h.
Records of annual physical inventory of all sealed source material of record.'
6.10.2.The following records:shall be retained'for the duration of the Unit Operating-License:
I a.
F.ecords and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.
b.
Records of new and irradiated fuel inventory, fuel transfers and assembly'burnup histories.
c.
Records of-radiation exposure for all individuals entering radiation control areas.
GE-STS 6-16
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fADMINISTRATIVE CONTROLS t
. RECORD RETENTION (Continued) d.
Records of gaseous and liquid radioactive material released to the environs.
e.
Records of transient or operational cycles for those unit components identified in Table 5.7.1-1.
f.
Records of reactor tests and s.ireriments.
g.
Records of training and qualification for current members of the unit
- staff, h.
Records of in-service inspections performed pursuant to these Technical Specifications.
i.
Records of Quality Assurance activities required by the Operational Quality Assurance Manual.
j.
Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
k.
Records of meetings of the (URG) and the (CNRAG).
t 1.
Records of the service lives of all hydraulic cnd mechanical snubbers listed on Tables 3.7.5-1 and 3.7.5-2 including the date at which the service life commences and associated installation and maintenance t
records.
6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared con-sistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA (OPTIONAL) 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100-mrem /hr but less than 1000 mrem /hr shall be barri-caded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit *.
Any individual or group of individuals permitted to enter such areas shall be pro-vided with or accompanied by one or more of the fnllowing:
~
a.
A radiation monitoring device which continuously indicates the radiation dose rate in the area.
l
- Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from.the RWP issuance requirement during the performance of their. assigned radiation p'rotection duties, provided they comply with approved radiation protection procedures for entry into.high radiation areas.
G:-STS.
6-17.
t ADMIN 15IRATIVE CONTROLS l
HIGH RADIATION AREA (OPTIONAL) (Continued) l l
b.
A radiation monitoring device which continuously integrates the radia-tion dose rate in the area and alarms when a praset integrated dose
~
l Jis received.
Entry into such areas with this monitoring device may L
be made after the dose rate level in the area has baen established and personnel have been made knowledgeable of them.
An individual qualified in radiation protection procedures who is c.
l equipped with a radiation dose rate monitoring device.
This indi-vidual shall be responsible for providing positive control over the i
l activities within.the area and shall perform periodic radiation l
surveillance at the frequency specified by the unit Health Physicist in the Radiation Work Permit.
i 6.12.2 The requirements of 6.12.1,.above, shall also apply to each high l
radiation area in which the intensity of radiation is greater than 1000 mrem /hr.
In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or the unit Health Physicist.
1 i
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I GE-STS 6-18
+
7 I
ALL STS-I SECTION 6.0 ADMINISTRATIVE CONTROLS hit are._ adukistrdive. Coptrols rolded fo rodid ical g
efflae#t ' tehcd sph ca tion wtda,. an_ to Lc Stad Wiki Sedien (o.D e.. % Gemeral E.lectric Stada.4 IEMud. S Edfi cdi o M (f )M6-Oln),,
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4 6.0 ADMINISTRATIVE CONTROLS L.
6.5.1 UNIT REVIEW GROUP (URG) i RESPONSIBILITIES
- 6. 5.1. 6 The URG shall be responsible for:
d k.
Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation,-recommendations and disposition of the corrective action to prevent recurrence to the (Superintendent of Power Plants) j and to the (Company Nuclear Review and Audit Group).
1.
Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems.
6.5.2 COMPANY NUCLEAR REVIEW AND AUDIT' GROUP (CNPAG)
AUDITS 6.5.2.8 Audits-of unit activities shall be performea under the cognizance of the (CNRAG). These audits shall encompass:
1.
The radiological environmental monitoring program and the results i
thereof at least once per 12 months, m.
The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.
2
- The PROCESS CONTROL PROGRAM and implementing procedures for solidifica-n.
tion of radioactive wastes at least once per 24 months, i
o.
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.
e 6.8 PROCEDURES 6.8.1,. Written procedures shall be established, implemented and maintained covering the activities referenced below:
g.
PROCESS CONTROL PROGRAM implementation.
h.
OFFSITE DOSE CALCULATION MANUAL implementation.
1.
Quality Assurance Program for effluent and environmental monitoring, using the guidance'in Regulatory Guide 4.15, December 1977.
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ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT,
lI 6.9.1.6 Routine radiological' environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.
The initial report shall be submitted prior to
_ May 1 of the year following initial _ criticality.
6.9.1.7 The' annual radiological environmental operating reports shall include summaries, interpretations, anJ an analysis of trends of the results of the radiological environmental surveillance ' activities for the report period, including a comparison with~preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.
The reports shall also include the results of land use censuses required _by Specification
~
3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report snall provide an analysis of the problem and a planned course of action to alleviate the problem.
i The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8,- December 1975 of all radiological environmental samples taken during the report period.
In the event that some results are. not available for inclusion with the report, the 4
report shall be submitted noting and explaining the reasons for the missing i
2 results. The missing data shall be submitted as soon as-possible in a
- supplementary report.
1 j
The reports'shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations 3
]
keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Comparison Program, 4
required by Specification 3.12.3.
3/
SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT
- 6. 9.1. 8 Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the l
first report shall_begin with the'date of initial criticality.
11 i'
3/
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 i
specify the releases of radioactive material from each unit.
i AL'. STS-I 6-2
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I ADMINISTRATIVE-CONTROLS 6.9.1.9 The. radioactive effluent release mports shall include a summary of the quantities of. radioactive. liquid and gaseous effluents and solid waste released from the' unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radio-s active Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a-quarterly basis following the format of Appendix 8 thereof.
The radioactive effluent release report to be submitted 60 days after January 1
~
i 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 i
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.
i joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaeous affluents released from the y
unit'or station during the previous calendar 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 (Figures 5.1-3 and 5.1-4) during the report period. - All assump-l tions used in making these assessments (i.e., specific activity,-exposure time 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 0FFSITE DOSE CALCULATION MANUAL (00CM).
The radioactive affluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to.the likely i
most exposed member of the-public from reactor releases and other nearby i
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 Operatin. Acceptable' methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1.
The radioactive effluents' relea'se shall include the following information for each type of solid waste shipped offsite during the report ~ period:
a.
Container volume, i
i b.
Total curie quantity (specifiy whether determined by measurement or
.1 estimate),
j c.
Principal: radionuclides ('pecify whether determined by measurement or estimate),
4 l
l ALL STS-I 6-3
o 4
4 ADMINISTRATIVE CONTROLS d.
Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),
i Type of container (e.g., LSA, Type A, Type B, Large Quantity), and e.
f.
Solidification 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) made during the reporting period.
MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submit +-ed on a monthly basis'to the Director, Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.
Any changes to the OFFSITE DOSE CALCULATION MANUAL shall 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 shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the (Unit Review Group).
PDOMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.11 j.
Offsite releases of radioactive materials in liquid and gaseous effluents which exceed the limits of Specification 3.11.1.1 or 3.11.2.1.
k.
Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the storage of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to-reduce the contents to within the specified limits.
f ALL STS-I 6-4
,- +
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ADMINISTRATIVE CONTROLS THIRTY DAY WRITTEN REPORTS 6.J.1.13 An unplanned offsite release of 1) more than 1 curie of radioactive e.
material in liquid effluents, 2)'more than 150 curies of noble gas in gaseous affluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. The report of an 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 unnlanned release.
3.
Actions taken to prevent recurrence.
4.
Consequences of the unplanned release, f.
Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.12-2 whea averaged over any calendar quarter sampling period.
6.10 RECORD RETENTION 6.10.2 1.
Records of analyses required by the radiciegical environmental monitoring program.
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ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior 13 f aplementation.
6.13.2 Licensee initiated changes to the PCP:
1.
Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
a.
Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; b.
A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and c.
Documentation of the fact that the change has been reviewed and found acceptable by the (URG).
2.
Shall become effective upon review and acceptance by the (URG).
6.14 0FFSITE DOSE CALCULATION MANUAL (00CM) 6.14.1 The 00CM shall be approved by the Commission prior to implementation.
6.14.2 Licensee initiated changes to the ODCM:
1.
Shall be submitted to the' Commission in the Monthly Operating Report-within 90 days of the date tha change (s) was made effective. This submittal shall contain:
a.
Sufficiently detailed information to totally support the rationale for the change without benefit of ' additional or supplemental information.
Information submitted should consist of a package of.those pages of the ODCM to be changed with each page numbered and provided with an approval and d3te box, together with i
appropriate analyses or evaluations justifying the change (s);
b.
A determination that the change will not reduce the accuracy or J
reliability of dose calculations or setpoint determinations; i
aM c.
Documentation of the fact that the change has been reviewed and found acceptable by the (URG).
2.
Shall become effective upon review and acceptance by the (URG).
i ALL STS-I 6-6
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l ADMINISTRATIVE CONTROLS
.1.
6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous and solid) l 6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):
1.
Shall be reported to the Commission in the Monthly Operating Report l
for the period in which the evaluation was reviewed by the (Unit Review Group). The discussion of each change shall contain:
a.
A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59; i
b.
Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; c.
A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; i
d.
An evaluation of the change which shows the predicted releases 4
of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license anplication and amendments thereto; e.
An evaluation of the change which shows the expected maximum exposures to individual in the ur, restricted area and to the general population that differ from those previously estimated in the license application and amendments thereto; f.
A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to.when the changes are to be made; g.
An estimate of the exposure to plant operating personnel as a result of the change; and h.
Documentation of the fact that the change was reviewed and j
found acceptable by the (URG).
2.
Shall become effective upon review and acceptance by the (URG).
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