Text

Application for Amend to Licenses DPR-58 & DPR-74,revising Radiological Effluent Tech Specs & Offsite Dose Calculation Manual to Incorporate Changes Agreed Upon in 820809 Conference Call,Per s Varga 820824 Ltr
	ML17334A428
Person / Time
Site:	Cook
Issue date:	09/17/1982
From:	Hering R; INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Harold Denton; Office of Nuclear Reactor Regulation
Shared Package
ML17319B516	List: ML17319B515; ML17334A428; ML20027B502;
References
	AEP:NRC:0055F, AEP:NRC:55F, NUDOCS 8209210136
	Download: ML17334A428 (188)
	v • d • e

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f REGULATORY ORMATION DISTRIBUTION SY 4 (RIDS)

ACCESSION NBR:8209210136 DOC ~ DATE; 82/09/17 NOTARIZFD: NO DOCKET FACIL:50 315 Donald C< Cook Nuclear Power Plant~ Unit l~ Indiana 8 05000315 50-316 Donald C, CooK Nuclear Power Planti Unit 2i Indiana 8 05000316 AUTH, NAME I AUTHOR AFF IL AT I ON HER ING g R, F ~ Indiana 8 Michigan Electric Co, REC IP ~ NAME RECIPIENT AFFILIATION DENTONgHBR ~ Office of Nuclear Reactor Regulations Director

SUBJECT:

Application for amend to Licenses DPR"58 8 DPR 70irevising radiological effluent Tech Specs L "Offsite Dose Calculation Manual" to incorporate changes agreed upon in 820809 conference calliper S Varga 820820 ltr<<

w Z ~g DISTRIBUTION CODE: AOOSS COP'IES TITLE: OR/Licensing Submittal; Appendix RECEIVED:LTR I

g NCL SIZE:,

NOTES:

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Dear Mr. Denton:

Please find enclosed a copy of the final version of the revised Radiological Effluent Technical- Specifications (RETS) for Unit 1 (Attachment 1) and Unit 2 (Attachment 2) of the Donald C. Cook Nuclear Plant and Offsite Dose Calculation Manual (ODCM) (Attachment 3) applicable to both Units. These documents are being submitted in accordance with Mr. S. Varga's letter of August 24, 1982. The RETS and ODCM have been revised to incorporate all the changes agreed upon in the conference call between representatives of the NRC, AEPSC, I&MECo. and Franklin Research Center on August 9,- 1982. These Specifications are to be incorporated into our Appendix "A" Technical Specifications. As agreed upon with your Staff, the requirements of these proposed Technical Specifications will not become effective until April 1, 1983.
At that time the existing Appendix "B" Technical Specifications, "Part 1 Radiological", will become null and void.
Included in Attachments 1 and 2 to this submittal is a cross reference of pages to be removed and inserted into the respective Appendix A Technical Specifications. Additionally, it should be noted that, for ease of processing, previous Technical Specification change submittals that affect the RETS replacement pages in Attachments 1 and 2 have also been included in this submittal. Attachment 4 lists the previous submittals and the affected pages. Only those current pages of Appendix A Technical Specifications which require changes have been submitted.
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Mr. H. R. Denton .. AEP:NRC: 0055F A minor editorial change has been introduced in Section 6.10, "RECORD RETENTION" of Attachment 1. The requirement to retain "Records of reactor tests and experiments" has been inserted in Section 6.10.2 and deleted from 6.10.1. The new requirement is more restrictive than the old one (i.e., retention for the life of the Plant instead of for five years) and will make Unit 1's requirements equal to those of Unit 2.
All new pages and changes to the existing Appendix "A" Technical Specifications have been so noted by a vertical bar on the right hand margin.
A unique situation exists in that new Radiation Monitoring System (RMS) equipment is currently being installed in both Units. Slight changes in the setpoint methodology for the new RMS monitors and in some of the ODCM tables describing the radiation monitors will have to be made. For some time, however, both the new and the old RMS will be in existence. The proposed Technical Specifications account for such intermediate situations. As new RMS equipment becomes operational,,
changes to the ODCM will be made as required by the RETS.
The final version of the revised RETS and ODCM has not been reviewed and approved by our safety review committees as of the submission date of this letter, They will be reviewed at meetings to be held prior to October 31, 1982. Please note, however, that our original RETS submittal, dated March 27, 1978 (AEP:NRC:0055) was reviewed both by the Plant's PNSRC and by AEPSC's NSDRC.
This application for a license amendment is considered to be a Class III Amendment as per 10 CFR 170. As required by Part 170, Subsection 22, a fee for $ 4,400.00 is required. Such a fee was transmitted with our letter to you, No. AEP:NRC:0055, dated March 27, 1978.
This document has been prepared following Corporate Procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.
Very truly yours, R. F. Hering Vice President
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Mr. H. R. Denton .AEP:NRC:0055F cc: John E. Dolan Columbus R. S. Hunter M. P. Alexich R. W. Jurgensen W. G. Smith, Jr. Bridgman R. C. Callen G. Charnoff Joe Williams, Jr.
J. G. Keppler - Regional Administrator U.S.
NRC Region III NRC Region III Inspector at Cook Plant Bridgman
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DEFINITIONS SECTION PAGE SOURCE CHECK ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 - 6 PROCESS. CONTROL PROGRAM (PCP) o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ 1 - 6 SOLIDIFICATION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 - 6 OFFSITE DOSE CALCULATION MANUAL (ODCM) ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 1 - 6 GASEOUS RADHASTE TREATMENT SYSTEM .............................. 1 - 6 VENTILATION EXHAUST TREATMENT SYSTEM ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 6 PURGE-PURGING ......................o.....o......oo............. 1-6 VENT ING ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6 MEMBER(S) OF THE PUBLIC ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-7 SITE BOUNDARY ~ o . o o o ~ ~ ~ ~ ~ o o o o o o . o o o o o o a ~ ~ o ~ o o e o, o ~ ~ o ~ o o o o o ~ o ~ ~ o o 1>>7 UNRESTRICTED AREA ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-7 D. C. COOK UNIT 1 I-A
INDEX LIMITING CONDITIONS FOR OPERATION .AND .SURVEILLANCE:REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3.4.2.1 Axial Flux Difference ............................... 3/4 2-1 3/4.2.2 Heat Flux Hot Channel Factor 3/4 2-5 3/4.2.3 Nuclear Enthalpy Hot Channel Factor . 3/4 2-9 3/4.2,4 Quadrant Power Tilt Ratio ...; ......,................ 3/4 2-11 3/4.2i5 DNB Parameters ........ ....... ......, .
~ ~ ~ ~ ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ 3/4 2-13 3/4.2.6 Axial Power Distribution ... . ..... . << .'. << ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-15 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION ................. 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION ~ ~ ~ ~ ~ ~ ~ ~ . ~ o ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ .
~ ~ ~ ~ ~ o ~ ~ ~ ~ 3/4 3-15 3/4.3. 3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation ................ 3/4 3-35 Movable Incore Detectors 3/4 3-39 Seismic Instrumentation ........,..................., 3/4 3-40 Meteorological Instrumentation ...... ......... ~ ~ . ~ ... 3/4 3-43 Remote Shutdown Instrumentation ............i........ 3/4 3-46 Fire Detection Instrumentation ........ ...... ~ ~ ~ ~ .. ~ . 3/4 3-51 Radioactive Liquid Effluent Instrumentation .. ~ ....., 3/4 3-57 Radioactive Gaseous Process and Effluent Monitoring Instrumentation .............,............ 3/4 3-62 3/4.4 REACTOR COOLANT SYSTEM 3/4o4o1 REACTOR COOLANT LOOPS Normal Operation ......,............................. 3/4 4-1 3/4.4.2 SAFETY VALVES SHUTDOWN ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-4 3/4.4,3 SAFETY VALVES. << OPERATING 3/4 4-5 3/4.4.4 PRESSURIZER ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-6 3/4.4.5 STEAM GENERATORS 3/4 4-7 D. C. COOK - UNIT 1 IV
LIMITING CONDITXONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTXON PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration ..........'....,......,.................... 3/4 ll-l Dose ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ r ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 ll 4 Liquid Waste Treatment 3/4 11-5 Liquid Holdup Tanks ....,.....,....,......,...... ~ ~ , ' , 3/4 11-6 3/4.11.2 GASEOUS EFFLUENTS Dose Rate ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 ~ ~ ~ ~ ~ ~ ~ 3/4 11-7 Dose Noble Gases ..................................... 3/4 11-10 Dose Radioiodines, Radioactive Material in Particulate Form, and Radionuclides other than Noble Gases 3/4 11-11 Gaseous Radwaste Treatment 3/4 11-12 Explosive Gas,Mixture ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-13 Gas Storage Tanks 3/4 11>>14 3/4.11.3 SOLID RADIOACTIVE WASTE . ~ ~ .. ~ ~ ~ ~ ~ ~ ~ o ~ o o o o o o ~ ~ ~ ~ ~ ~ . o 3/4 11-15 3/4.11.4 TOTAL DOSE o...woo..o...oooo..a.coo...g....o...o.... 3/4 11-17 3/4.12 RADIOLOGICAL'ENVIRONMENTAL MONITORING 3/4,12.1 MONITORING PROGRAM ......,...o...ao.oo.o.o.o.o...... 3/4 12'1 3/4.12.2 LAND USE CENSUS ...................,................ 3/4 12-9 3/4.12.3 INTERLABORATORY COMPARXSON PROGRAM ......,.....,.... 3/4 12-10
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D. C. COOK UNIT 1
ZNDEX BASES SECTION 3/4. 11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS .. ~ o ~ ~ . ~ ~ ~ ~ ~ oo o ~ .o o ~ o ~ ~ . ~ o ~ . ~ ~ . ~ ~ ~ . B 3/4 11-1 3/4.11,2 GASEOUS EFFLUENTS,........ ~.........,.......,...... B 3/4 11-2 3/4.11.3 SOLID RADIOACTIVE WASTE ............-,...;.........., B 3/4 11-5 3/4.11.4 TOTAL DOSE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 1 1 5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM ............,.....,....... << .. << ~ B 3/4 12-1 3/4.12.2 LAND USE CENSUS ..................................., B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ................. B 3/4 12-1 D. C. COOK UNIT 1
INDEX OES IGN FEATURES SECTION PAGE 5.1 SITE E xclusion Area................,.............................. 5-1 Low Population 7one.................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-1 Site Boundary for Gaseous and Liquid Effluents........,..... 5-1
5. 2 CONTAINMENT C onfiouration................................................
f-Design Pressure and Temperature...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 5-4 P enetrations'......................... ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-4 5.3 REACTOR CORE F uel Assemblies................................. . " ..... ~ ~ ~ ~ ~ 5-4 Control Rod Assemblies........,..........,................... 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature.............................. 5-4 V 01 Ume ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5-5 5.5 EMERGENCY CORE COOLING SYSTEMS............................... 5-5 5.6. FUEL STORAGE C t rl tlcali 1 ty.... ~... ~... .. .. ~....,
~ ~ ~ ~ ~ . ~ . ~ . ~ . ~ ~ ~ ~ ~, ~ ~ ~ ~ ~, 5-5 0 rainaget ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ p ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 ~ ~ ~ ~ ~ ~ 5-6 C apacity.................................................... 5-6 5.7 SEISMIC CLASSIFICATION.......,...,.........,...,,...,...,... 5-6 5.9 MET OROLOGICAL 70'rJER LOCATION......................,........
5.9 COMP-NENT O'YCLE OR TRANSIENT LIMIT..........;............,... 5-o D. C. COOK - UNIT 1 XVI
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o INDEX ADt>IHISTPATIYE, CONTROLS SECTION PAGE 6.1 RESPONSI B IL ITY..............................................
6o2 ORGANIZATION 0 f t ffsite..... ~ .. ~ ~ . ~ .. ~ ~ . ~ ~ - ~ ~ ~ ~ ~ ~ ~ ~ ~ - ~ ~ ~ " ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ 6-1 F
l 'L Staff.................;;. ..
acility ~ ~ ~ . ~ ..-. ~ ~ ~ ~ " ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-1
'o3 FACILITY STAFF QUALIFICATIOHS..............................., 6-5 6 o4 TRAIHI oG o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
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6. 5 REVIEW AND AUDIT 6.5.1 PLAti'T NUCLEAR SAFETY REVIEi< COMMITTEE F unction........................-.'........................ G-G ]
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o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-6 A lternates................................................ 6-6 et ecting Frequency..............,........................- ". 6-6 Q uOrum o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-6 Responsibilities.... o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-7 Au IrlhlOr ty 1 o ~ ~ ~ ~ ~ ~ ~ ~ ~, ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ 6-8 R ecords...............'..................................... 6>>8 6.5.2 NUCLEAR SAFETY AHD DESIGN REVIEW'COMMITTEE unc ion............................o................-....
OKDONpo 1 tl On ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
6>> p Alternates............o..................
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0. C. C00'K - UNIT l XVII
MINISTRATIVE CONTROLS SECTION PAGE Consultants 6-9 Meeting Frequency 6-9 Q UO ZCR ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
Reviev ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-10 Audits ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-11 Aut110rity ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ 6 12 Records ooooooooo ~ ooooooooooooooo ~ ooooooooooooooooo ~ ~ o ~ ~ ooo 6-12 6 ' REPORTABLE OCCURRENCE ACTION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 12 6 ' SAFETY LIMIT VIOLATION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 13 6o8 PROCEDURES o..ooo.oo..o..oo.oo..o...oooo..o.o...ooo..oooooo 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES ~ ~ ~ ~ .. ,.
~ 6-14 6.9.2 SPECIAL REPORTS ...o.......o..o...o.o.o..o....o..oo. 6-22 6.10 RECORDS RETENTION oo.o..o....o.oooooo...oooo.oo...ooooo.... 6-23 6.11 RADIATION PROTECTION 'PROGRAM ~ ~ ...,......... ..............
~ 6-24 6.12 HIGH"RADIATION AREA o ~ ~ ~ ooo ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ o.,oo ~ ~ ~ oo ~ o ~ ~ o ~ o ~ ~ 6-24
6. 13 ENVIRONMENTAL UALIFICATION ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ 6~24 6.14 PROCESS CONTROL PROGRAM .....o.........o..oo............... 6-25 6.15 OFFSITE DOSE CALCULATION MANUAL . ~ oo ~ ~ ~ ~ ~ ~ .o ~ oo ~ ~ ~ o ~ ~ ~ ~ ~ o ~ . 6-25 6.16 HAJOR CHANGES TO RADIOACTIVE'WASTE I
'TREATMENT SYSTEMS ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ oo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-26 D. C. COOK - UNIT 1 XVIII
DEFINITIONS PRESSURE 8OUNDARY LEAKAGE
'.16 PRESSURE 80UNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through.a non-isolabl e fault in a Reactor Coolant System component body, pipe wall -or vessel wall.
CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump'eals.
OUADRANT POMER TILT RATIO 1.18 QUADRANT POMER TILT RATIO 'shall be the ratio of the maximum upper.
excore detector calibrated output to the average of the upper excor e detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore det ctor calibrat d outputs, whichever is greater. Mich one excore detector inoperable, .he remaining three detectors shall be used for computing the average.
DOSE EOUIYALENT I-131 l.l9 .00SE E(UIVALEHT I-131 shall be that concentration of I-131 (pCi/gram) which alone ~ould produce the same thyroid dose as the quantity and iso-topic mixture of I-131, I-132, I-133, 7-134, 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 Diswnc Factors for Power and iest Reactor Sites, or in NRC Regulatory Guide 1.109 Rev. 1, October 1977.
STAGGERED TEST 8ASIS 1.20 A STAGGERED TEST SASIS shall consist of:
a. A "est schedule for n systems, subsystems, trains or o her designated components obtained by dividing the specified test interval into n equal subint rvals,
b. The testing of one sys.em, subsystem, train or other designa.ed componen a" the beginning of each subinterval.
Q. C. COOK -. UNIT 1
1.0 DEFINITIONS SOURCE CHECK GZI G G*.G,GGG I 111 ".1 1 11:
when the Channel sensor is exposed to a radioactive source.
ll'esponse PROCESS CONTROL PROGRAM PC?)
1. 8 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analysis, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes'ill be accomplished in such a way as to assure compliance with 10 CFR 20, 10 CFR 71, Federal and State regulations and other requirements governing the shipment and disposal of. radioactive waste.
SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of radioactive liquid, resin and sludge wastes" rom liquid systems into a form that meets shipping and burial site requirements.
OFFSITE DOSE CALCULATION MANUAL ODCM)
.30 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and para-meters used in the calculation of offsite doses due to radiactive gaseous and liquid efi'luents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints and the conduct of environmental radiological monitoring program.
GASEOUS RADMASTE TR~D&ii ENT SYSTBI 1.31 A GASEOUS RAOWASTE TREATMENT SYSTB1 is any system designed and installed to reduce radioactive aaseous effluents by collecting primary coolant system off-aases from the primary system and providing for delay or holdup for the purpose of reducing- the total radioactivity- prior to release to the environment.
VEtlTILATION EXHAUST TREAW>>PilT SYSTEM 1.32 A VENTILATION EXHAUST TREATMENT SYST&u is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulat s from the gaseous exhaust str am prior to the release to the envii onment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREA7NENT SYSTEM! comoonents.
PURGE-PURGING G T II GIIRGIGG gas =rom a confinement I;1 to maiintain 11 d 1 df 1 Gf 1 emperature, pressure, humidity, concentration or other operating condition, in such a manner that placement air or gas is required to purify the confinement.
r-VBlTING 1.34 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, oressur , humidity, concentration or other operating condition, in such a manner that replac ment air or gas is not provided or required during VENTING. Vent, used in sys .em names, does not imply a VENTING process.
D. C. COOK - UNIT 1 1-6
0EFINITIONS i4lPIBER S OF THE PUBLIC
.35 NEiSER S OF THE PUBLIC shall include all persohs who are not occupationally associated uith the plant. This category does not include employees of the utility, its contractors or its vendors.
Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions gf the site for recreational, occupational or other purposes not associated with the plant.
A SITE BOUNDARY 1.36 The SITE BOUNDARY- shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.
UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for :
purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential qua'rters or industrial, commercial, institutional and/or recreational purposes.
D. C. COOK - UNIT 1 1-7
TASLE i OPERATIONAL MODES REACTI YITY :o RATED COOLANT iXOOE CONDITION 5'VERAGE THERMAL POWER* TEMPERATURE L. POWER OPERATION > 0.99 5N > 350'F.
2. STARTVP > 0.99 > 350'F
3. HOT STANDSY < 0.99 > 350'F
4. HOT SHUTDOWN < 0.99 350 F >Tavg
> 200'F
5. COLO SHUTDOWN < 0.99 ( 200'F
6. REFUELING~ < 0.95 < 140'F
" Excluding decay heat. ~
~ Reactor vessel head unbol'ted 'or removed and fuel'n the vessel. ~
D. C. COOK - UNIT 1 1-8
TABLE I
l. 2 FREQUENCY NOTATION NOTATION FREQUENCY AT LEAST ONCE PER 12 HOURS AT LEAST ONCE PER 24 HOURS AT LEAST ONCE PER 7 DAYS AT LEAST ONCE PER 31 DAYS AT LEAST ONCE PER 92 DAYS SA AT LEAST ONCE PER 184 DAYS AT LEAST ONCE PER 549 DAYS S/U PRIOR TO EACH REACTOR START-UP COMPLETED PRIOR TO EACH RELEASE N.A. NOT APPLICABLE D. C. COOK UNIT 1 1-9
INSTRUMENTATION RADIOACTIYE LI U!D EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3;9 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3. 11. 1.1 are not exceeded.
APPLICABILITY: As shown in Table 3.3-12.
ACTION:
a. Nith a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of 3.11. 1. 1 are met, without delay suspend the release .of radioactive liquid effluents monitored by the affected channel, reset, or declare the channel inoperable.
b. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the applicable ACTION shown in Table 3.3-12.
c. The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE RE UIREMENTS 4.3.3.9. 1 The setpoints shall be determined in accordance with methodology as described in the ODCM and shall be recorded.
4.3.3.9.2 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8.
D. C. COOK
~ ~
- UNIT 3. 3/4 3-sj
TABLE 3.3-12 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument ~0erabl e A 1 i cabi 1 i t Action
1. Gross Radioactivity Monitors Providing Automatic Release Termination
a. Liquid Radwaste At times of rel ease Effluent Line
b. Steam Generator At times of release 24 Blowdown Line
c. Steam Generator At times of release Blowdown Treatment Effluent
2. Gross Radioactivity Monitors Not Providing Automatic Release Termination
a. Service Hater System At all t'imes 25 Effluent Line
3. Continuous Composite Sampler Flow Monitor
a. Turbine Building Sump At all times 25 Effluent Line It
4. Flow Rate Measurement Devices
a. Liquid Radwaste Line (1) At times of release 26
b. Discharge Pipes* (1) At al 1 times NA
c. Steam Generator Blowdown Treatment Effluent (1) At times of release 26
Pump curves and valve settings may be utilized to estimate flow; in such cases, Action Statement 26 is not applicable.
D. C. COOK - UNIT 1 3/4 3-58
TABLE 3.3-12 (Continued)
TABLE NOTATION Action 23 With the number of'hannels OPERABl E less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed for up to 30 days, provided that prior to initiating a release:
1. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1:. and;
2. ,At least two technically qualified members of the Facility Staff independently verify the discharge valving. Otherwise, suspend release of radioactive effluents via this pathway.
Action 24 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for gross radioactivjty (beta or gamma) at a limit of detection of at least 10 qci/gram:
1. At least once once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is >0. 01 ~ci/gram DOSE EQUIVALENT I-131.
2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific a'ctivity of the secondary coolant is <0.01 uci/gram DOSE EQUIVALENT I-131 .
Action 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for gross radioactivity (bet~ or gamma) at a lower limit of detection of at least 10 pci/ml.
Action 26 With the number of Channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.
D. C. COOK - UNIT I 3/4 3-59
TABLE 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK: CHECK CALIBRATION TEST
l. Gross Beta or Gamma Radioactivity Monitors providing alarm and automatic isolation
~
a. Liquid Radwaste R(3) 0(I)
Effluent Line
5. Steam Generator R(3) Q(i)
Blowdown Effluent Line O'Jf
c. Steam Generator R(3)
Blowdown Treatment Effluent Line
2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm But Not Providing Automatic Isolation
a. Service Mater R(3)
System Effluent Line
3. Continuous Composite Samplers
a. Turbine Building N/A N/A N/A Sump Effluent Line
4. Flow Rate Monitors *
a. Liquid Radwaste O(4) N/A Effluent *
b. Steam Generator O(4) N/A N/A N/A Blowdown Treatment Line
" During Releases Via This Pathway
0. C. COOK - UNIT I 3/4 3-60
TABLE 4.3-8 (Cont)
TABLE NOTATION The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:
1. Instrument indicates measured levels above the ala'rm/
trip setpoint.
Z. Circuit failure.*
3. Instrument indicates a downscale failure.*
~ 4. Instrument control not set in operating mode.*
.(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions ex 1 sts l
1. Instrument indicates measured levels above the alarm setpoint.
2. Circuit failure.
3. Instrument indicates a downscale failure.
4. Instrument controls not set in operating mode.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more sources with traceability back to the National Bureau of Standards. These sources shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.
(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic or batch re'leases are made.
Instrument indicates, but does not provide for automatic isolation.
+ As equipment becomes operational.
D. C. COOK - UNIT 1 3/4 3-61
Instrumentation Radioactive Gaseous Process and Effluent Monitorina Instrumentation Limitin Condition for 0 eration 3.3.3.10 The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of 3. 11. 2. 1 are not exceeded.
A~)i b1 n
i: A 0 i T bl E.3-13.
Action:
a. With a radioactive gaseous process or effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a .value which will ensure that the limits of 3.11.2.1 are met~ without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, reset, or declare the channel inoperable.
b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.
C. The provisions of Specification 3.0.3, 3.0,4 and 6.9.1.13 are not applicable.
Surveillance Re uirements 4.3.3.10.1 The setpoints shall be determined in accordance with methodology as described in the ODCM and shall be recorded.
4.3.3.10.2 Each radioactive gaseous process or effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAI TEST operations at the frequencies shown in Table 4.3-9.
This surveillance requirement does not apply to the Waste Gas Holdup System Hydrogen and Oxygen Monitors, as their setpoints are not addressed in the ODCM.
D. C. COOK - UNIT 1 3/4 3-62
TABLE 3.3-13 Radioactive Gaseous Effluent Monitorin Instrumentation Minimum Channels Instrument ~0erab1e A licabilit Maste Gas Holdup System Explosive Gas Monitoring System
a. Hydrogen Monitor (1) 30
b. Oxygen Monitor (2) 29
2. Condenser Evacuation System
a. Noble Gas Activity b.
Monitor Flow Rate Monitor (1)
(1)
~'ction28 27
3. Auxi1 iary Bui1 ding Yentilation System
a. Noble Gas Activity (1)
Monitor
b. Iodine Sampler (1) 32 Cartridge
c. Particulate (1)
Sampler Filter
d. Effluent System (1) 27 Flow Rate Measuring Device
e. Sampler Flow Rate 27 Measuring Device Containment Purge System ***
a. Noble Gas Activity (1) *4'**1 31 Monitor
b. Particulate Sampler (1) 32 Maste Gas Holdup System
a. Noble Gas Activity 33 Monitor Providing Alarm and Termination of Gas Decay Tank Releases
6. Gland Seal Exhaust
a. Noble Gas Activity Monitor 28
b. Flow Rate Monitor D. C. COOK - UNIT 1 3/4 3-63
TABLE 3.3-13 (Cont)
At all times
'o (treatment for primary system gases During waste gas holdup system o p eration mp e con ainment atmosphere not containment ur e.
cont ainmentt, activity.
termination of purge on high con During releases via this pathway 1
For purge ur ose pur e purposes only, see Technical Specifications 3.3.3.1 ~ ~ ~ ) 3 ~ 4~6 o 1 and 3.9.9 for other requirements.
For gas decay tank reIeases only see 3 f a dd'r i tional requirements.
COOK << UNIT 1 3/4 3-64
0. C.
TABLE 3.3-13 (Cont)
TABLE NOTATION Action 27 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef luent releases via this pathway may continue for up to 30 days provided the flow rate is es imated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
Action 28 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
Action 29 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of this system may continue for up to 14 days. With 2 channels in-operable,,operation of this system may continue for up to 14 days, provided grab samples are taken and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Action 30 Mith the number of channels OPERABLE less than required b'y the Minimum Channels OPERABLE requirement, operation of this system may continue for up to 14 days, provided grab samoles are taken and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Action 31 With the number of channels OPERABLE less than reouired'by the Minimum Channels OPERABLE requirements, immediately suspend PURGING of radioactive effluents via this pathway.
Action 32 Mith the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 4. 11-2.
Action 33 With the number of channels OPERABLE less than r quired by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be r leased to the environment for up to 14 days provided that prior to initiating the release:
a. At least two independent samples of the tank's contents are analyz d and,
b. At least two technically qualified members of the Facility Staff independently ve'rify the r lease rate calculations and discharge valve lineups; otherwise, suspend release of radioactive effluents via this pathway.
0. C. COOK - UNIT 1 3/4 3-65
TABLE 4.3-9 Radioactive Gaseous Effluent Monitorin Instrumentation Surveillance Requirements Channel Channel Source Channel Functional Instrument Check Check Calibration Test:.
I. Waste Gas Holdup System Explosive Gas Monitoring System
a. Hydrogen Q*** NA a(3)
Monitor
b. Oxygen 0~* a(4)
Monitor
c. Oxygen Q*** Q(4)
Monitor(Alt.)
Z. Condenser Evacuation System
a. Noble Gas Activity 0*
R(Z)
Monitor
b. System Effluent 0"' "'
Flow Rate .
3. Auxil iary Building ation System 'entil
a. Noble Gas Activity 0* R(Z)
Monitor
b. Iodine Sampler W* NA NA NA
c. Particulate Sampler W* NA NA NA
d. System Effluent 0" NA R Q Flow Rate Measure-ment Device
e. Sampler Flow Rate 0" Measurement Device
4. Containment Purge System
a. Noble Gas Activity 0~ R(Z) C(5)
Monitor
b. Particulate Sampler W~ NA NA
5. Waste Gas Holdup System
a. Noble Gas Activity P** R(2) O(5)
Monitor Providing Alarm 8 Termination of Gas Decay Tank Releases
0. C. COOK - UNIT j. 3/4 3-66
TABLE 4.3-9 (Cont)
6. Gland Seal Exhaust
a. Noble Gas Activity D** R(S)
Monitor
b. System Effluent NA Flow Rate
At all times During Release Via This Pathway
~* During l<aste Gas Holdup System Operation (Treatment for Primary System Offgases)
D. C.
~ ~ COOK. - UNIT I 3/4 3-67
TABLE 4.3-9 (Cont)
TABLE NOTATION The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
1. Instrument indicates measured levels above the alarm setpoint.
2 Circuit failure.
lpga' Instrument indicates a downscale failui e.
Instrument controls not set in operate mode.
2) The initial CHANNEL CALIBRATION shall be performed using one or more sources with traceability back to the National Bureau of Standards. These sources shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial, calibration may be used.
3) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
l. One volume percent hydrogen, balance nitrogen, and
2. Four volume percent hydrogen, balance nitrogen.
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1. One volume percent oxygen, balance nitrogen, and
2. Four volume percent oxygen, balance nitrogen.
The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:
2
l. Instrument indicates measured levels above the alarm/trip setpoint.
Circuit failure.+
3 Instrument indicates a downscale failure.*
Instrument controls not set in operate mode.*
Instrument indicates, but does not provide automatic isolation.
~ As equipment becomes operational.
D. C. COOK - UNI'T 1 3/4 3-68
3 4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 Li uid Effluents Concentration
~Li i i di i 0 3.11.1.1 The concentration of radioactive material released at any time from the site to unrestricted areas (see Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases.
For dissol~ed or entrained noble gases, the concentra tion shall be limited to 2 N 10 vci/ml total activity.
~A)i ill: A Action:
With the concentration of radioactive material released from the site ex-ceeding the above limits, without delay restore the concentration to wi.thin the above limits.
SURVEILLANCE RE UIREi~1ENTS
4. 11. 1 '. 1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4. 11-1.
4. 11. 1. 1.2 The result of radioactive -analysis shall be used in accordance with the methods of the ODChl to assure that all concentrations at the point of release are maintained within the limits of Specification 3. 11. 1. 1.
D. C. COOK - UNIT 1 3/4 ll-l
TABLE 4.11-1 Radio'active Liquid Waste Sampling and Analysis Program Lower Limit a Minimum Type Of of Detection Liquid Release Sampling Analysis Activity (LLD)
T e Fre uenc Fre uenc Anal sis >ici/ml A. Batch Waste Release Tanks P P. Principal 5X10 Gamma e
tters
'mi Each Batch Each Batch I-131 1 xl 0 Dissolved lxlo and Entrain-ed Gases (Gamma One B'atch/M Emitters)
P H-3 'X10 Each Batch Composite Gross Alpha 1X10-P Q b Sr-89,Sr-90 5X10
-6 Each Batch Composite re- 1X10 B. Plant Continuous Daily Principal 5X10 Releases d t Gamma Emitters Composite I-131 lxlo Dissolved 1X10 and Entrain-Grab Sample ed Gases (Gamma Emitters)
Daily M osite H-3 1X10 Com Gross Al ha 1X10 Dai ly Q b Composite Sr-89, Sr-90 SX10 Fe-55 lxlo-6 D. C. COOK - UNIT 1 3/4 11-2
TABLE 4.11-1 (Cont)
JABLE NOTATION
a. The lower limit of detection (LLD) is defined in Table Notation
a. of Table 4.12-1 of Specification 4.12.1.1.
b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is re-presentative of the liquids released.
c. A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analysis, each batch shall be isolated and re-circulated to ensure thorough mixing.
d. A continuous release. is the discharge of liquid waste of a non-discrete volume; e.g. from a volume of system that has an input flow during the continuous release.
e. The principal gamma emitters for which the LLO specification applies exclusively are the following radionuclides: Mn-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, together with the above nuclides, shall also be identified and reported.
D. C. COOK - UNI'T 1 3/4 11-3
RADIOACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to an individual from radioactive material in liquid effluents released to unrestricted areas (see Figure 5.1-3) shall be limited:
a. During'ny calendar quarter to <1.5 mrem to the total body and to <5mrem to any organ, and
b. During any calendar year to < 3 mrem to the total body and to < 10 mrem to any organ.
APPLICABILITY: At all times.
ACTEON:
a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of.-the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause (s) for exceed-ing the limit(s) and defines the corrective actions taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be within the above limits.
This Special Report shall also include (I) the results of radiological analyses of the drinking water source, and (2) the radiological impacts on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Hater Act. (Applicable only body.')
if drinking water supply is taken from the receiving water
b. The provisions of Specification 3.0.3, 3:0.4 and 6.9.1.13 are not appl i cab 1 e.
SURVEILLANCE RE UIREMENTS 4.11.1.2 Dose Calculations.:Cumulative dose contributions from I iquid effluents shall be determined in accordance with the Offsite Dose Calcu-lation Manual (ODCM) at least once per 31 days.
D. C. COOK - UNIT 1 3/4 11-4
Radioactive Effluents Li ui d Was te Trea tmen t Limitin Condition For 0 eration 3.11;"I.3 The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the pro-jected doses due to the liquid effluent from the site (s'ee Figure 5.1-3) when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.
~At i bill A Action:
4
a. ':lith radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare.and submit to the Commission within 30 days pursuant to Specification 6;9.2 a Special Report which includes the following information:
l. Identification of the inoperable equipment or subsystems and the reason for inoperability,
2. Action(s) taken to restore the inoperable equipment to operabIe status, and
3. Summary description. of action(s) taken to prevent recurrence.
b. The provisions of Specification 3.0.3, 3'.0..4 and'.9: 1'.'l3: are not..
applicable.
Surveillance Re uirements 4.11,1.3 Doses due to liquid releases to UNRESTRICTED AREAS shal.l be pro-jected'at least once per 31 days, in accordance with the QDCH, whenever liquid releases are being made without being processed by the liquid rad-waste treatment system.
D. C. CQQK - UNIT 1 3/4 II-6
Radioactive Effluents r
Li uid Holdu Tanks*
Limitin Condition For 0 eration
3. 11.1.4 The quanti ty of radioactive material contained in each of the following tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.
a. Outside temporary tanks.
A Action:
a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limi t, without delay suspend all addi tions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.
b. The provisions of Specification 3.0.3, 3.0.4 and. 6; 9.:.1;13 .are. not applicable.
Surveillance Re uirements 4.11.1.4 The quantity of radioactive materi'al contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.
Tanks included in this Specification are those outdoor tanks tha't are not surrounded by liners, dikes, or walls capable. of holding the tank contents and that do not have, tank overflows and sur-rounding area drains connected to the liquid radwaste treatment system.
D. C. COOK - UNIT 1 3/4 11-6
Radioactive Effluents 3/4.11.2 Gaseous Effluents Dose Rate timitin Condition For 0 eration 3.11.2. 1 The dose rate due to radioactive materials released in gaseous effluents from the site (See Figure 5.1-3) shall be limited to the following:
a. For noble gases: < 500 mrem/yr to the total body and < 3000 mrem/yr to the skin, and
b. For all radioiodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than 8 days: < 1500 mrem/yr to any organ.
~All ill: A d1.
Action:
lilith the dose rate(s) exceeding the above limits, without delay decrease the release rate to within the above limit(s).
Surveillance Re uirements 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM.
4.11 '.1.2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM by ob-taining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.
D. C. COOK - UNIT 1 3/4 11-7
~ ~
~ ~
~ 0 '
~
~ ~
~ ll ~
~ ~ ~
~ ~
~ ~
~ N
~ ~ ~
TA8LE 4.11-2 (cont)
TABLE NOTATION
a. The lower limit of detection (LLD) is defined in Table Notation a.
of Table 4.12-1 of Specification 4.12.1.1.
b. Analyses shall also be performed following any operational occurrence which has altered the mixture of radionucl ides as indicated by RCS analysis. (ie., start-up.)
C. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing. Analyses shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 .days following each shutdown, startup or similar operational occurrence which lead to significant increases or decreases in radioiodine in the Reactor Coolant System. >lhen samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor of 10.
d. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or 'dose rate calculation made in accordance with Specification 3, 11.2. 1, 3.11.2.2, and 3.11:2.3.
e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133M, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported, Other peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and reported.
D. C. COOK - UNIT 1 3/4 11-9
RADIOACTIVE EFFLUENTS DOSE NGBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose in UNRESTRICTED AREAS due to noble gases released in gaseous effluents shall be limited to the following:
a. During any calendar quarter., to < 5 mrad for gamma radiation and < 10 mrad for beta radiation;
b. During any calendar year, to < 10 mrad for gamma radiation and
< 20 mrad for beta radiation.
APPLICABILITY: At al 1 times.
ACTION:
a. Mith the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, 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 reduce the releases and the proposed corrective actions to be taken to assure that subsequent re-
'leases will be within the above limits.
b. The provisions of Specification 3. 0.3, 3.0.4 and 6. 9. 1. 13 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for the total time period sha be determined in accordance with the Offsite Dose Calculation Manual (ODCN) at least once every 31 days.
D. C. COOK - UNIT 1 3/4 11-10
RADIOACTIVE EFFLUENTS DOSE RADIOIODINES RADIOACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER 'THAN 'NOBL'E 'GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to A MEMBER OF THE PUBLIC from radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days in gaseous effluents released to unrestricted areas shall be limited to the following:
a. During any calendar quarter to ~ 7.5 mrem to any organ; b, During any calendar year to ~ 15 mrem to any organ.
APPLICABILITY: At all times.
ACZZON:
Mith the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits, 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 and defines the corrective actions taken to reduce the releases and the proposed corrective action to be taken to assure that subsequent releases will be within the above limits.
b. The provisions of Specification 3.0,3, 3.0.4, and 6.9,1.13 are not applicable.
SURVEILLANCE 'RE UIREMENTS 4,11,2.3'DOSE CALCULATIONS Cumulative dose contributions for the total time period shall be determined in accordance with the ODCM at least once every 31 days.
D. C. COOK << UNIT 1 3/4 11-11
Radioactive Ef f1 uents Gaseous Radwaste Treatment Limi tin Condition For Ooeration 3.11.2.4 The gaseous radwasta treatment system and the ventilation exhaust treatment system shall be used to reduce the radioactive materials in gaseous waste prior to their discharge when the proje'ctad gaseous effluent air doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1.3) when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The ventilation exhaust treatment system shall be "used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1-3) when averaged over 31 days would exceed 0.3 mram to any organ.
~A)i bi1i: A i
Acti an:
a. Mith gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which in-cludes the following information:
l. Identification of the inoperable equipment or subsys .ams and the reason for inoperability.
2. Action(s) taken to restore the inoperable equipment to operable s atus.
b. The provisions of Specification 3.0.3, 3.0.4 and 6.9. 1.13 are not applicable.
Surveillance Reouirenents 4.11,2. 4 Qoses due io gaseous releases to UNRESTRICTEQ AREAS shall be pro'ected at least once per 31 days in accordance with the OQCN, whenever the gaseous waste treatment system or ventilation exhaust trea~nent system is not operational.
D. C. COOK - UNIT 1 3/4 11-12
RADIOACTIVE EFFLUEiNTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR'OPERATION 3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to < 2% by volume if the hydrogen in the system is > 4% by volume.
APPLICABILITY: At al 1 times.
ACTION:
a. With the concentration of oxygen in the waste gas holdup system
> 2% by volume but < 4% by volume and containing > 4% hydrogen, restore the concentration of oxygen to < 2% or reduce the hydrogen concentration to < 4% within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .
b. With the concentration of oxygen in the waste gas holdup system or tank > 4% by volume and > 4% hydrogen by volume without de'lay suspend all additions of waste gases to the system or tank and reduce the concentration of oxygen to < 2% or the concentration of hydrogen to
< 4% witnin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months in the system or tank.
c. The provisions of Specification 3.O.3, 3.O.4 and 6.9.l.l3 are not applicable.
SERI'EIIIRREE RRIEIIIREIIERIS
4. 11.2.5 The concentration of oxygen in the waste gas holdup system shall be determined to within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3. 10.
D. C. COOK - UNIT 3. 3/4 11-13
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RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADIOACTIVE WASTE LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be used as applicable in accordance with a PROCESS CONTROL PROGRAM for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.
APPLICABILiTY: At al 1 times.
ACTION:
a ~ With the packaging requirements of 10 CFR Part 20 and/or 10 CFR Part 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site.
b. With the solid radwaste system inoperable for more than 31 days, in lieu of any other report required by Specification .
6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:
l. Identification of the inoperable equipment or subsystems and the reason for inoperability,
2. Action(s) taken to restore the inoperable equipment to operable status,
3. A description of the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and
4. Summary description of action(s) taken to prevent a recurrence.
c. The provisions of Specification 3.0.3, 3.0.4 and 6.9. 1.13 are not applicable...
D. C. COOK - UNIT 3, 3/4 11-15
SOLID RADIOACTIVE WASTE SURVEILLANCE RE UIREMENTS 4.11.3.1 The solid radwaste system shall be demonstrated operable at least once per 92 days by:
a. Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or
b. Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.
4.11.3.2 THE. PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type 'of wet radioactive waste (e.g. filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).
a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with .the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.
b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.14 to assure SOLIDIFICATION of subsequent batches of waste.
0. C. COOK - UNIT 1 3/4 11-16
RADIOACTIVE EFFLUENTS 3/4 11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to < 25 mrem to the total body or any organ (except the thyroid, which is lim7ted to < 75 mrem) over a period of 12 consecutive months.
APPLICABILITY: At all times.
ACTION:
a. Mith the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b, 3. 11.2.3.a, or
3. 11.2.3.b, in lieu of any other report required by Specification 6.9. 1, prepare and submit a Special Report to the Director-, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Mashington, D. C. 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3. 11.4. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 12 consecutive, month period that includes the release(s) covered by this report.
exceeds the limits of Specification 3.11.4, and ifIf the estimated'ose(s) the release condi'tion resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified infor-mation of 6 190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as addressed in other sections of this Technical Specification.
h
b. The provisions of Specification 3.0.3, 3.0.4 and 6.9.l.13 are not applicable.
SURVEILLANCE RE UIREMENTS 4.11.4 g ~h Dose Calculations
4. 11.1.2, 4.11.2."2, and Cumulative dose contributions from d d dd 4,11.2.3 and with the d
ODCH.
dh d 1 iquid and
0. C. COOK - UNIT 1 3/4 11-17
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'RADIOLOGICAL EN V IRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION CONTINUED)'.
With milk or fresh leafy vegetable samples unavailable from any of the sample locations required by Table 3. 12-1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause of the unavaila-bility of samples and identifies locations for obtaining replace-ment samples. The locations from which samples were unavailable may then be deleted from Table 3. 12-1 provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations, available.
if
d. The provisions of Specification 3.0.3, 3.0.4 and -6.9;1.13 are not applicable.
SURYEILLANCE RE UIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3. 12-1 from the locations given in the table and figures in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and'.12-1.
D. C. COOK - UNIT 1 3/4 12-2
TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Type of Frequency and/or Sam les Sam le Locations Collection Fre uenc of Anal sis
1. Airborne
a. Radioiodine Al-A6 (Site) Continuous operation Radioiodine cani s and Particulates of sampler with Analyze New~ Buffalo, Sample Collection Weekly for I-131 South Bend, as required by Dowagiac, and Oust Loading But At Particulate sampl~
Coloma are Least once Per.7 Days Gross Beta Rad-Background ioactivity follow Filter Changea , ~
posite (by loca-tion) for gamma isotopic quarterl~
Z. Direct Radiation a) Tl-T9 (Site) At least once per Gamma Dose. At 92 days Least Once Per b) New Buffalo 92 days South Bend Oowagiac Coloma c) 10 TLD Monitor Locations in the Five Mile Radius
3. Waterborne
a. Surface Ll , L2 , L3 Composite Sample Gamma Isotopic Over One-Month Period Analysis monthly .
Composite For tritium analysis quarterly.
b. Ground Wl-W 7 tluarterly Gamma Isotopic and Tritium analys quarterly .
c. Drinking St. Joseph Composite Sample Gross Beta and Lake Township Collected over a Gamma Isotopic New Buffalo Period of < 31 days Ana lysi s of each Composite" Sample composite sample.
Over a 2 week Tritium Analysis Period if I-131 of composite Analysis is Performed. quarterly.
I-131 analysis on ,
each composite whe~
collecting the dose calculatet omposite samples shall be collected by an aliquot for the consumptio~
at intervals not exceeding 24 hours. of the water is greater than 1 mrer D.. C. COOK - UNIT 1 3/4 12-3
TABLE 3.12-1 (Cont)
d. Sediment from LZ, L3 2/year Gamma Isotopic Shoreline I Analyses Semi-Annual ly.
4. Ingestion a.Niglk Stevensville At least once per Gamma Isotopic Bridgman 15 days when animals and I-131 Analysi Galien are on Pasture. At of Each Sample.
Oowagiac Least Once Per 31 cmouth Bend Days at Other Times.
o~'g
b. Fish Plant Si te 2lyear Gamma I sotopi c Off-Site Analysis on Edibli Portion.
c. Food Products Plant Site At time of Harvest Gamma Isotopic Off-Site (approx. One Sample of Each Analysis on Edibl~
.20 mi) of the Following Portion.
Classes of Food Products
1. Grapes Plant Site At time of Harvest Gamma Isot One sample of Broad Analysis Leaf Vegetation a
Particulate sample filters should be analyzed for gross beta 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampIing to allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 times the yearly mean of control sampIes for any medium, gamma isotopic analysis should be performed on the individual samples.
D. C. COOK - UNIT 3. 3/4 12-4
TA8Lc 3 12-2 Reporting Levels For Radioactivity Concentrations In Environmental Samples Food Mater Airborne Particulate Fi h Products Analysis (pci/1) or Gases (pci/m ) (pciAg,wet) (pci/1) (pcijkg,
! >tet) l H-3 2 X 10 Nn-54 1X10 3X10 4
Fe-59 4 X 10 1X10 3 4 Co-58 1X10 3 X 10 CO-60 3 X 10 1X'l0 2
Zn-65 3X10 2 X 10 2
Zr-Nb-95 4 X 10 I-131 0.9 1 x lo 3
s-134 30 10 1 xlo 60 1X10 3
, CB-137 50 20 2 X 10 ~
70 2X10 2
, Ba-La-140 2 X 10 3 X 10 D. C. COOK -.
UNIT 3. 3/4 12-5
TABLE 4.12-1 iMaximum Values For The Lower Limits of Detection (LLD)
Airborne Particulate Analysis Mater or Fish hiilk Food Products Sediments (pci/1) Gas 3
(pci/kg) (pci'/1) (pci/kg,wet) (pci/kg,dr (pci/m ) wet
-2 Gross Beta 1 x 10 H-3 2000 Hn-54 15 130 Fe-59 260 Co-58, 60 130 Zn-65 30 260 Zr-95 30 15
'b-95
-2 I-131 7X10 60
-2 Cs-134 15 6X10 130 15 60 150
-2 Cs-137 18 6X10 150 18 60 180 Ba-140 60 60 La-140 15 15 D. C. COOK - UNIT j. 3/4 12-6 0
TABLE 4.12-1 (Cont)
TABLE NOTATION
a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
LLD =
4.66 sb E
' '.22 ~
Y ~
exp (-Rat) where LLD is the "a priori" lower .limit of detection as defined above (as pci per unit mass or volume),
sb is the standard deviation of the background counting rate or of toe counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency(as counts per transformation),
V is the sample size (in units of mass or volume),
2.22 is the number of transformation per minute per picocurie, Y is the fractional radiochemical yield (when applicable),
X is the radioactive decay constant for the particular radionuclide and
~t is the elapsed time between sample collection (or end of the sample collection period) and time of counting. I The volume of Sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appro-pr iate) rather than on an unverified theoretically predicted variance.
In calculating the. LLD for a radionuclide determined by gamma-ray spec-trometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., potassium - 40 in milk samples).
0. C. COOK - UNIT 1 3/4 12-7
~
Table 4.12-1 (Cont)
Table Notation Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of inter-ferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contribution factors will be identified and described in the Annual Radiological Environmental Operating Report.
b. LLD for drinking water.
c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identified and reported.
D. C. Coo< - UNn Z 3/4 12-8
RADIOLOGICAL ENVIRONMENTAL t10NITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall ide'ntify the loca-tion of the nearest milk animal, the 'nearest residence and the nearest garden* of greater than 500 square feet producing fresh leafy vegetables in each of the . 9 land covering meterological sectors within a distance of five miles.
APPl ICABILITY: At all times.
ACTION:
a. With a land use census identifying a location(s) which yields a calculated dose or dose commitment greater than the values
currently being calculated in Specification 4.11.2.3, pre-pare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the net location(s).
b. With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exoosure pathway) 20 percent areater than at a location from which samples are currently being obtained in accordance with Specification 3.t2.1, prepare and submit to the Comm>ssion within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new location. The new location shall be added.to the radiological environmental monitoring program within 30 days, if possible.
The sampling location having the lowest calculated dose or 'aose commitment, (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted.
c. The orovisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicab'Ie.
SURVEILLANCE RE UIREMENTS 4.12.2. The land use census shall be conducted at least once per 12 months between the dates of June I and October I, by door-to-door survey, aerial survey, or by consulting local agricuIture authorities.
8 g t gt g yg g y g tt g g the direction sector with the highest D/g in lieu of the garden census.
D. C. COOK - UNIT 1 3/4 12-9
Radiolo ical Environmental Monitorin 3 4 12.3 Interlaborator Com arison Pro ram Limiting Condition For Operation 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which hasobeen approved by the Commission.
~AII I III: A Action:
a. Mith analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.
b. The provisions of Specification 3. 0.3, 3. 0,.4 and 6.9. 1;13 are not applicable.
Surveillance Re uirements 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the OGCM (or
'articipants in the EPA crosscheck program shall provide the EPA program code designation for the unit) shall be included in the Annual Radiological Environmental Operating Report. c
0. C. COOK - UNIT j. 3/4 12-10
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3/4. I I I RAQ I OACT V E EF FLUEN7S 8ASES 3/4.11.1 LIOUIG EFFLUENTS 3/4.11.1.1 CONCENTRATION. This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas uill be less than the concentration levels speci-fied in 10 CFR Part 20, Appendix 8, Table II. This limitation 'provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures within (I} the Section II.A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR Par. 20.106(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its HPC in air (subnersion) was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protec-tion (ICRP) Publication 2.
3/4.II.I.2 OOSE. This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operaton implenents the guides set forth in Section II.A of Appendix I. Tne ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section I'/.A of Appendix I to assure that the releases of radioactivematerial in liquid effluents will be kept "as low as is reasonably achievable." Also, for fresh water sites wi.h drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40: CFR I41.
The dose calculations in the ODQI imolement-the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be show'n by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be subs,antially underestimated. The equations specified in the OQCN for calculating the doses due to the ac.ual release rates of radioac.ive materials in liquid effluents uilj be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Ooses to Man from Routine Releases of Reac.or Effluents
=or the Purpose of Fvaluating Compliance with 10 CFR Part 50, Appendix I,"
Revision I, Cc.ober 1977, and Regulatory Guide 1.113, "Estimating Aouatic Qisper sion of Effluents from Accidental and Routine Reac.or Releases for the Purpose of Implementing Appendix I," April 1977. HUREG-0133 provides methods or dose calculations consistent uith Regulatory Guide 1.109 and 1.113.
D. C. COOK - UNI'T j. 8 3/a 11-1
RAG IOACTIYE EFFLiJEt'TS BASES This specification applies to the release of liquid effluents from each reactor at the site. The liquid effluents from the shared system are proportioned among the units sharing the system.
3/4.11.1.3 LIQUID MASTE TREATMENT. The operability of the liquid radwaste treatment system ensures that this system wil I be available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the. appropriate portions. of this system be used when speci-fied provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This speci-fication implenents the requirements of 10 CFR Part 50.36a, General Design Criteria Section'11.1of the Final Safety. Analysis Report for the Donald C.
Cook Nuclear Plant, and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate po~tions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFP, Part 50, for liquid effluents.
3/4.11.1.4 LIQUID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in theevent of an uncontrolled release of the tanks'ontents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.
3/4.11.2. GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE. This specification is provided to ensure that the dose rate any any time at the SITE BOUNDARY from gaseous effluents frowy all units on the site will be within the annual dose limits of 10 CFR Part 20 for unres.ricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, o annual average concentrations exceeding the limits specified in Appendix 8, Table II of 10 CFR Part 20 (10 CFR Part 20.106 (b) ).
For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the a&ospheric di> usion factor above that for the si e boundary.
The specified release rate limits restrict, at all times, the corre-sponding garma and beta dose rates above background to an individual at or beyond the site boundary to < (500) mrem/year to the total body or to <(3000) mrem/year to the skin. These release rate limits aIso restrict, at all times, the corresponding thyroid dose rate above back-ground to an infant via the cow-milk-infant pathway to < 1500 mr em/year for the nearest cow to the plant.,
This specification applies to the release of gaseous effluents =rom all reactors at the site. The gaseous effluents from the shared system are proportioned among the units sharing that systen.
D. C. COOK - UNIT 1 B 3/4 11-2
RADIOACTIVE EFFLUENTS BASES 3/4.11.2.2 DOSE, NOBLE 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 Condition for. Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendix' to assure that the releases of radio-active material in gaseous effluents will be kept "as low as is reasonanle achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous 'effluents will be 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 19?? and Regulatory Guide l.ill, "Methods for Estimating Atmos-pheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Hater-Cooled Reactors," Revision 1, July 1977. The ODCM equations.-provided for determining the air doses at the site boundary will be based upon the historical average atmospherical conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.
3/4.11.2.3 DOSE, RADIOIODINES, RADIOACTIVE MATERIAL IN PARTICULATE'FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES. This specification is provided to implement 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 impl'ement the guides set forth in Section IV.A of Appendix I to assuage that the releases of radioactive materials in gaseous effluents will be= kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data sich that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods approved by the NRC for calculating the doses due to the actual release rates of'he subject materials are required to be 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 D. C. COOK - UNIT 1 B 3/4 11-3
RADIOACTIVE EFFLUENTS BASES Gaseous Effluents in Routine Releases from Light-Mater-Cooled Reactors, Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.
The release rate specifications for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependant on the existing, radionuclide pathways to man, in the unrestricted area.
The pathways which are examined in the development of these calculations are: 1) individual inhalation of'irborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat pro-ducing 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 MASTE TREATMENT The operability of the gaseous radwaste treatment system and the ventila-tion exhaust treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion Section ll. 1 of the Final Safety Analysis Report for The Donald C. Cook Nuclear Plant and design objective Section II.D of Appendix I. to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
3 4 11.2.5 Explosive Gas thixture This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treat-ment system is maintained below the flammability limits of hydrogen and oxygen mixtures. Haintaining the: concentration of hydrogen or oxygen below their flammability limits provides that the releases of radioactive materials will be controlled ir, conformance with the requirements of the General Design Criterion specified in Section 11. 1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.
3/4 11.2.6 Gas Stora e Tanks Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an in-dividual at the nearest site boundary will not exceed 0.5 rem. This is consistant with Standard Review Plan 15.7.1. "Maste Gas System Failure."
D. C, COOK. - UNIT 1 B 3/4 11-4
RADIOACTI VE EF FLUENTS.
BASES 3/4.11.3 SOLID RADIOACTIVE WASTE The operability of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to shipment offsite, This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion specified in Section 11.1 -of the Final Safety Analysis. Report for the Donald C. Cook Nuclear Plant. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limit-ed to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing time.
3/4.11.4 TOTAL DOSE The specification is provided to meet the dose limitations qf 40 CFR 190, The specification requires the preparation and submittal of a Special Report svhenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I'or sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40'FR 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 limitations 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 to 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 provision of 40 CFR 190.11 is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle.
D. C. COOK - UNI'T 1 8 3/4 11-5
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12. 1 MONITORING PROGRAM The radiological monitoring program required by this specification provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radio'iogical effluent monitoring program by verifying that the measurable concentration of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.
The detection capabilities required by Table 4.12-1 are the state-of-the art for routine environmental measurements in industrial laboratories.
It should be recognized that the LLD is defined an an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.
Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable.
In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.
3 4.12.2 LAND USE CENSUS This specification i's provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (25 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.
To determine this minimum garden size, the following assumptions were used, 1) that 20% of the garden was used for growing broad leaf vegetation (i.e. similiar to lettuce and cabbage), and 2) a vegetation field of 2 kg/square meter .
3 4.12.3 INTERLABORATORY CGMPAPISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive mater ial in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.
D. C. COOK - UNIT 1 B 3/4 12-1
5.0 DESIGN FEATURES 5.1 SITE Exclusion Area 5.1.1 The exclusion area shall be shown in Figure 5.1-1.
Low Po ulation Zone 5.1.2 The low population zone shall be shown in Figure 5.1-2.
Site Boundar For Gaseous and Li uid Effluents 5.1.3 The site boundary for gaseous and liquid effluents shall be shown in Figure 5.1-3.
5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
a 0 Nominal inside diameter = 115 feet.
b. Nominal inside height = 160 feet.*
c~ Minimum thickness of concrete walls = 3'".
d. Minimum thickness of concrete roof = 2'".
e. Minimum thickness of conrete floor pad = 10 feet.
f. Nominal thickness of steel liner, side and dome = 3/8 inches.
g. Nominal thickness of steel liner, bottom = 1/4 inch.
h. Net free volume = 1.24 x 10 cubic feet.
From grade (Elev. 608') to inside of dome.
D. C. COOK - UNIT 1 5-1
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FIGURE .5.1-3 SITE BOUNDARY FOR GASEOUS NlD LIQUID EFFLUENTS
r a ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Protection Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in Plant design, and response and analysis of the Plant for transients and accidents .
6.3,2 Until the newly appointed Operations Superintendent obtains a Senior Reactor Operator's License, all of his licensed functions will be performed by a full time assistant who holds a current Senior Reactor Operator's License.
6 '4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Plant Manager and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976, D. C. COOK - UNIT 1 6-5
ADMINISTRATIYE CONTROLS
6. 5 REVIEW AND AUD IT 6.5.1 PLANT NUCLEAR SAFETY REYIEM COMMITTEE PNSRC)
FUNCTION 6.5.1.1 The PNSRC shall function to advise the Plant Manager on
all matters reIated to nuclear safety.
COMPOSITION 6.5.1. 2 The PNSRC shall be composed of the:
Chairman Plant Manager or designated al ternate Member: Assistant Plant Managers Member: Operations Superintendent Member: Technical Superintendent Member: Maintenance Superintendent Member: Control and Instrument Supervisor
~
Member: Nuclear/Computer Engineering Supervisor Member: Plant Chemical Supervisor Member: Performance Supervising Engineer Member: Plant Radiation Protection Supervisor Member: Shift Supervisor Member: Environmental Coordinator ALTERNATES 6.5.1.3 All alternate members shall, be appointed in writing by the PNSRC Chairman to serve on a temporary basis; howeve~, no more than two alternates shall participate as voting members in PNSRC activi ties at any one time.
MEETING FRE UENCY 6.5.1.4 The PNSRC shall meet at least once per calendar month and as convened by the PNSRC Chairman or his designated alternate.
UORUM 6.5.1.5 A quorum of the PNSRC shall consist of the Chairman or his designated alternate and four members including alternates.
D. C. COOK - UNIT 1 6-6
ADMINISTRATI>lE CONTROLS RESPONSIBILITIES 6.5.1.6 The PNSRC shall be responsible for:
Review of 1) all procedures required by Specifica'tion 6.8 and changes thereto, 2) any other proposed procedures or changes thereto as determined by the Plant Manager 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 propose'd changes or modifications to plant systems or equipment that affect nuclear safety.
e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairman of the NSDRC.
Review of those REPORTABLE OCCURENCES requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission.
Review of facility operations to detect potential safety hazards.
Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the NSDRC.
Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.
Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.
k. Review of every unolanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the NSDRC.
Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems.
D. C. COOK - UNIT j. 6-7
ADMINISTRATIVE CONTROLS AUTHORITY
6. 5.1. 7 The PNSRC shall:
a. Recommend to the Plant Manager written approval or 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 unreviewed safety question.
c ~ Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the NSDRC of disagreement between the PNSRC and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The PNSRC shall maintain written minutes of each meeting and copies shall be provided to the Chairman of the NSDRC.
6.5.2 NUCLEAR SAFETY AND DESIGN REVIEW COMMITTEE NSDRC FUNCTION 6.5.2.1 The NSDRC 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 mechanical and electrical engineering quality assurance practices D. C. COOK - UNIT 1 6-8 0
ADMINISTRATIVE CONTROLS .
COMPOSITION 6.5.2,2 The NSDRC shall be composed of the'.
Chairman: Assistant Vice President, Nuclear Engineering Member: Vice Chairman, Engineering and Construction Member: President and Chief Operating Officer of I&MECo Member: Executive Vice President, Construction & New York Engineering Member: Vice President, Mechanical Engineering Member: Vice President, Electrical Engineering Member: Vice President, Engineering Administration Member: Assistant Vice President, Design Division Member: Assistant Vice President, Environmental Engineering Division Member: Plant Manager, D. C. Cook Plant Member: Manager, Nuclear Safety and Licensing Section Alternate: Assistant Chief Mechanical Engineer Alternate: Assistant, Plant Manager, D. C. Cook Plant Alternate: Executive Assistant to the President of I&MECo Alternate: Assistant Division Manager, Nuclear Engineering ALTERNATES 6,5.2.3 All alternate members shall be appointed in writing by the NSDRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NSDRC activities at any one time.
CONSULTANTS 6.5.2,4 Consultants shall be utilized as determined by the NSDRC Chairman to pr'ovide expert advice to the NSDRC.
6.5.2.5 The NSDRC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.
QUORUM 6.5.2.6 A quorum of NSDRC shall consist of the Chairman or his designated alternate and at least six NSDRC members including alternates. No more than a minority of the quorum sha11 have line responsibility for operation of the facility, D. C. COOK UNIT 1 6-9
r e AEI'lINIS ) RATI'IE CONTROLS MD)TS 6.5.2.8 Audits of facility act'vities shall be per ormed under the cognizance of the HSDRC. Tnese audits shall encompass:
a. The conformance of facility operation to provisions contained within the Technical Soecification and applicable license conditions at least once per 12 months.
The performance, training and qualifications of the entire facility staff at least once per 12 months.
C. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.
Tne performanc Program to me onc~ per 24 months.
t:of ac;ivi.ies he c."it r-a of required Appendix by the "8",
Ouali-y "ssur nc 10 CFR =0, at leas-
e. Tne Facility Emergency Plan and implementing procedures at least onc per 24 months.
f. The Facility Security Plan and implementing procedures at least once per 24 months.
g Any other area i of facility operation considered appropriate by the )VSDRC.
The Facility Fire Protection Program and implemen0ing procedures at leas once per 24 months.
An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either qualified.offsite lic nsee or an, outside fir protection firm. personnel An inspection and audit of the fire protection and loss pre-vention program shall be performed by a qualified ou.side fire consultant a least once per 36 months.
The radiological environment=-1 monitoring program and the results thereof at least once per 12 months.
The OFFSITE DOSE CALCULATION VA)VUAL and implementing orocedures at least once pe. 24 months.
D. C. COOK - UNIT 1 6-11
ADMINISTRATIVE CONTROLS m, The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months.
n. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 1.21, Rev. 1, June 1974 and Regulatory Guide 4.1, Rev. 1, April 1975 at least once per 12 months.
AUTHORITY 6.5.2.9 The NSDRC shall report to and advise the Vice Chairman, Engineering and Construction, AEPSC, on those areas of responsibility specif ied in Sections 6.5.2. 7 and 6.5.2.8.
RECORDS 6.5,2.10 Records of the NSDRC activities shall be prepared, approved and distributed as indicated below: . /
a. Minutes of each NSDRC meeting shall be prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days- following each meeting.
b. Reports of reviews encompassed by'Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Vice Chairman, .
Engineering and Construction, AEPSC, within 14 days following completion of the review.
c. Audit reports encompassed by Section 6.5.2.8 above,'hall be forwarded to the Vice Chairman, Engineering and Construction, AEPSC, and to the management positions responsible for the areas audited within 30 days after completion of the audit.
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 houi notification to the Commission, shall be reviewed by PNSRC and submitted to the NSDRC Chairman.
D. C. COOK - UNIT 1 6-12
ADMINISTRATIVE CONTROLS 6.7 . SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
a. The facility shall be placed in at least HOT STANDBY within once hour.
b. The Safety Limit Violation shall be reported to the Commission and to the Chairman of the NSDRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PNSRC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or
'structures, and (3) corrective action taken to prevent recurrence.
d. The Safety Limit Violation Report shall be submitted to the Commission, the Chairman of the NSDRC, and the Executive Vice, President-Construction and New York Engineering, AEPSC, within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
a~ The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.
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.
g. PROCESS CONTROL PROGRAM implementation.
h. OFFSITE DOSE CALCULATION MANUAL implementation.
Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 1.21, Rev. 1, June 1974 and Regulatory Guide 4.1, Rev. 1, April 1975.
6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PNSRC and approved by the Plant Manager prior'o implementation and reviewed periodically as set forth in administrative procedures.
D. C. COOK - UNIT 1 6-13
ADHINISTRATI VE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 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 Hay 1 of each year.
6.9.1.7 The annual radiological environmental operating reports shall include summaries, interpretations, and statistical evaluation 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 surveil-lance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the land use censuses required by Specification 3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the repor t shall provide an analysis of the problem and a planned course of action to alleviate the problem.
/
The annual radiological environmental operating reports shall include summarized and tabu'lated results in the format of 3.12-2 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 report shall be submitted noting and explaining tne reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following: a summary description of the radiological environmental monitoring program including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equip-ment used; a map of all sample locations keyed to a table giving distances and directions from one reactor; the result of land use census required by the Specification 3.12.2; and the results of participation in the Inter'laboratory Comparison Program. required by Specification 3.12.3.
P 3l 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 specifiy the releases of radioactive material for each unit.
0. C. COOK - UNIT 1 6-16
ADMINISTRATIVE CONTROLS SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.S 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.
6.9.1.9 The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units as outlined in Regulatory Guide 1.21, "Measuring, Evaluating and Reporting in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-llater Cooled Nuclear Power Plants," with data summarized on a quarterly basis following the format of Appendix B, thereof.
The radioactive effluent release report to be submitted 60 days after January 1 and July 1 of each year shall include a quarterly summary of hourly meteorogical data collected during the reporting period. This summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction and atmospheric stability. The report submitted 60 days after January 1 shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. These reports shall include as assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figure 5.1-3) during the reporting period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. The meteorological condi tions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampIing frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM).
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 specify the releases of radioactive material from each unit.
D. C. COOK - UNIT 1 6-17
ADMINSTRATIVE CONTROLS The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. l.
The radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report period:
a. Volume (cubic meters),
b. Total curie quantity (specify whether determined by measurement or estimate),
c. Principal radionuclides (specify whether determined by measurement or estimate),
d. Type of waste (e.g;, spent resin, compacted dry waste, evaporator bottoms),
e. Type of container (e.g., LSA, Type A, Type B, Large guantity),
and
f. Solidification agent (e.g., cement).
The radioactive effluent release report shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluent on a quarterly basis.
The radioactive effluent release reports shall include any change to the PROCESS CONTROL PROGRN (PCP) and the OFFSITE DOSE CALCULATION NNUAL (ODCH) made during the reporting period.
i~10NTHLY REACTOR OPERATING REPORT 6.9."1.10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office Of Management and Program Analysis, U.S. Nuclear Regulatory Coamission, 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.
REPORTABLE OCCURRENCES 6.9.1.11 The REPORTABLE OCCURRENCES of Specification 6.9.1.12 and 6.9.1.13 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution 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.
D. C. COOK - UNIT 1 6-18
ADMIMISTRATIVE CONTROLS PROMPT NOTIFICATION WITH WRITTEN FOLLOMUP 6.9.1.12 The types of events listed below shall be reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone and confirmed by telegraph, mailgram or facsimile transmission to the Oirector of the Regiona1 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 circumstances 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 monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure .to complete the required protective function.
b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condi tion 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 operation greater than or equal to 1/ak/k; a calculated reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technical specification; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more xnan 0.5Xak/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 requirements of system(s) used to cope with accidents analyzed in <he SAR.
f. Personnel error or procedural inadequancy. which prevents or could prevent, by itself, the fulfillment of the functional reouirements of systems required to cope with accidents analyzed in the SAR.
D. C. COOK - UNIT 1 6-19
ADMINISTRATIVE CONTROLS
g. Conditions arising from natural or man-made events that, as a direct result of the event require unit shutdown, operation of safety systems, or 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 analvsis in the safety analysis report or technical specification 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..
J ~ Offsite releases of radioactive material 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 with-in the specified limits.
D. C. COOK - UNI'T 1 6-20
ADHINISTRATI>/E CONTROLS HIRTY DAY WRITTEN REPORTS 6.9.1.13 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occurence 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 explana-tion of the circumstances surrounding the event.
a. Reactor protection system or engineered safety feature in-strument 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 re-quired by a limiting condition for operation.
c. Observed inadequacies in the implementation of administrative or procedural 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 than those specified in 6.9.1.12.c above designed to contain radioactive'aterial resulting from the fission process.
e. An unplanned offsite release of 1) more than 1 curie of radio-active material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information:
l. A description of the event and equipment involved.
2. Causes(s) for the unplanned release.
3. Actions taken to prevent recurrence.
4. Consequences of the unplanned release.
Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.12-2 when averaged over any calendar quarter sampling period. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
D. C. COOK - UNIT 1 6-21
MINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 ~ Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional OFfice within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference Specifications:
a. Inservice Inspection Program Review, Specification 4.4.10.
b. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
c. Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3 ~
d. Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.
e. Seismic event analysis, Specification 4.3.3.3.2.
f. Sealed Source leakage on excess of limits, Specification 4.7.7.1.3.
g. Fire Detection Instrumentation, Specification 3.3.3.7.'.
Fire Suppression Systems, Specifications 3.7.9.1, 3.7.9.2, 3.7.9.3 and 3.7.9.4 D. C. COOK - UNIT 1 6-22
MINISTRATIVE CONTROLS 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
a. Records and logs of unit operation covering 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 sealed source leak tests and results.
go Records of annual physical inventory of all sealed source material on record.
6.10.2 The following r'ecords shall be retained for the duration of the Facility Operating License:
a. Records 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.
d. Records of gaseous and liquid radioactive material released to the environs.
e. Records of radioactive shipments.
f. Records of transient or operational cycles for those facility components identified in Table 5.9-1.
go Records of training and qualification for current members Plant staff.
of'he
h. Records of in-service inspections performed pursuant to these Technical Specif ications.
Records of Quality Assurance activities required by the QA Manual.
Records of reviews performed for changes made to procedures or equipment or review of tests and experiments pursuant to 10 CFR 50.59.
k. Records of meetings of the PNSRC and the NSDRC.
l. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13,
m. Records of reactor tests and experiments.
D. C. COOK UNIT 1 6-23
ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent 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
6. 12.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20:
a ~ A High Radiation Area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation llork Permit and any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.
b. A High Radiation Area in which the intensity of radiation is greater than 1000 mrem/hr shall be subject to the provisions of 6.12.l.a above, and in addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the adminstrative control of the Shift Operating Engineer on duty.
6. 13 ENVIRONMENTAL UALI F I CAT I ON 6.13.1 By no later than June 30, 1982, all safety-related electrical equipment in the facility shall be qualified in accordance with the providions of: Division of Operating Reactors "Guidelines for Evaluating Environmental gualification of Class 1E Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental gualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License No. DPR-74, dated October 24, 1980.
6. 13.2 By no later than December 1, 1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.
D. C. COOK - UNIT 1 6-24
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ADMINISTRATIVE CONTROLS Qi 2. Shall become effective upon reyiew and acceptance by the PNSRC.'.15.3 Commission initiated changes:
1. Shall be determined by the PNSRC to be applicable to the facility after consideration of facility design.
2. The 1icensee shall provide the Commission with written noti-fication of- their determination of applicability including any necessary revisions to reflect facility design.
6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Li uid, Gaseous and Solid 6.16.1 Licensee initiated major changes 'to the radioactive waste systems (liquid, gaseous and solid):
1. Shall be reported to the Commission in the Annual Operating Report for the period in which the evaluation was reviewed by the (PNSRC)
The discussions 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;
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;
d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;
e. An evaluation of the change which shows the expected maximum exposure to individuals in the unrestricted 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; D. C.~ COOK
~
- UNIT I 6-26
ADMINISTRATIVE CONTROLS
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 found acceptable by the PNSRC.
Z. Shall become effective upon review and acceptance by the PNSRC.
6;16.2 Commission initiated changes:
1. The applicability of 'the change to the facility shall be determined by the (PNSRC) after consideration of the facility design.
2.. The licensee shall provide the Commission with written notification of its determination of applicability including any necessary revisions to re f1 ect faci 1 i ty design.
0. C. COOK - UNIT I 6-27
ATTACHMENT 2 TO AEP:NRC:0055F D. C. COOK - UNIT 2
ATTACHMENT 2 TO AEP:NRC:0055F LIST OF CHANGES TO APPENDIX A TECHNICAL SPECIFICATION D. C. COOK - UNIT 2 DPR-74 REMOVE PAGES INSERT PAGES I-A IV IV IX-A XIV-A XV XV XVI XVI XVII XVII 1-4 1-4 1-6 to 1-8 1-6 to 1-10 3/4 3-53 to 3/4 3-54 3/4 3-53 to 3/4 3-66 3/4 ll-l to 3/4 11-17 3/4 12-1 to 3/4 12-10 B 3/4 3-3 B 3/4 3-3 B 3/4 11-1 to B 3/4 11-5 B 3/4 12-1 5-1 5-1 5-9 6-5 to 6-9 6-5 to 6-9 6-11 to 6-13 6-11 to 6-13 6-16 to 6-21 6-16 to 6-28
INDEX DEFINITIONS SECTION PAGE SOURCE CHECK ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 - 6 PROCESS CONTROL PROGRAM (PCP) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 6 SOLIDIFICATION ...................... o o . o o... o ...............'.. 1-7 o OFFSITE DOSE CALCULATION MANUAL (ODCM) .......... .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ 1 -7 GASEOUS RADWASTE TREATMENT SYSTEM ...,...,...................... 1 - 7 VENTILATION EXHAUST TREATMENT SYSTEM ....,............... ~... .. ~ 1 - 7 PURGE PURGING ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ 1-7 VENTING ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-7 MEMBER(S) OF THE PUBLIC 1-8 SITE BOUNDARY ............o.....o << ..... << ...a.....o..i...oaeoo. 1-8 UNRESTRICTED AREA ...ooo ~ ~ ~ .oooo oooaooo
~ ~ o ~ ~ oo,oooo aooo oeoeoa, 1-8
~ ~
D. C. COOK - UNIT 2 I -A
LIMITING CONDITIONS .FOR OPERATION AND SURVEILLANCE .REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 Axial Flux Difference 3/4 2-1 3/4.2.2 Heat Flux Hot Channel Factor 3/4 2-5 3/4.2,3 RCS Flow Rate and R ....,......,.....,,...,,....,..., 3/4 2-9 3/4.2,4 Quadrant Power Tilt Ratio 3/4 2-13 3.4.2.5 DNB Parameters 3/4 2-15 3/4.2.6 Axial Power Distribution ........................... 3/4 2-17 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION .......,........ 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION .................................... 3/4 3-14 3/4 .3. 3 MONITORING INSTRl2KNTATION Radiation Monitoring Instrumentation . ~.....,.....,. 3-34 Movable Incore Detectors 3/4 3-38 Meteorological Instrumentation .....,....... ~....... 3/4 3-39 Remote Shutdown Instrumentation ..........,......... 3/4 3-42 Post-Accident Instrumentation . ~......... . .. . .. 3/4 ~ ~ ~ ~ ~ 3-45 Axial Power Distribution Monitoring System ......,,. 3/4 3-48 Fire Detection Instrumentation ......,......
fluent'nstrumentation.'/4 3/4 3-50 Radioactive Liquid Ef 3/4 3-53 Radioactive Gaseous~Process'nd Effluent Monitoring Instrumentation' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3"58 3/4.3.4 TURBINE OVERSPEED PROTECTION ~ .. ~ ~ . ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 3/4 3-65 3/4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 REACTOR COOLANT LOOPS Normal OPeration ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-1 3/4.4.2 SAFETY VALVES SHUTDOWN 3/4 4-4 3/4.4.3 SAFETY VALVES OPERATING 3/4 4-5 3/4 i4 ~ 4 PRESSURIZER ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-6 3/4 4
~ ~ 5 STEAM GENERATORS ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-7 D. C. COOK - UNIT 2 IV
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4. 11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration ,.............................,...,...,... 3/4 11-1 Dose ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ qo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11 4 Liquid Waste Treatment ...,............,....,..., .
(
~ ~ ~ ~ ' 3/4 11-5 Liquid Holdup Tanks ...,......,.....,....,.......,.... ~, 3/4 11-6 3/4.11.2 GASEOUS EFFLUENTS Dose Rate ~ 000100100000 ~ 0000100000,000000001 ~ 444000040000 3/4 11<<7 Dose - Noble Gases 3/4 11-10 Dose Radioiodines, Radioactive Material in Particulate Form, and Radionuclides other than N oble Gases .....oooeooo.........oo.o.mao ~ ~ ~ 4 ~ bt ~ ~ ~ I, ~ ~ ~ .3/4 11-11 Gaseous Radwaste Treatment 3/4 11-12 Explosive Gas Mixture ....,.........,..........,....,... 3/4 11-13 Gas Storage Tanks 3/4 11-14 3/4,11,3 I
SOLID RADIOACTIVE WASTE .. ~ ..... ..............
~ ~ ~ ~ ~ . 3/4 11-15 3/4.11.4 TOTAL DOSE o.ooooooo.woo.oeooeooo'oo.ooooaooooooooo.o 3/4 11-17 3/4.12'ADIOLOGICAL ENVIROhtENTAL MONITORING 3/4.12.1 MONITORXNG PROGRAM .......................,...... .. 3/4 12>>1 3/4.12.2 LAND USE CENSUS ....,.................,,.........,.. 3/4 12-9 3/4.12.3 XNTERLABORATORY COMPARISON PROGRAM ................. 3/4 12-10 D. C. COOK - UNIT 2 IX-A
0 INDEX BASES SECTION PAGE 3/4. 11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS .................................. B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS ......,.......................... B 3/4 11-2 3/4.11.3 SOLID RADIOACTIVE HASTE ..................... ~ ~ ~ ~ ~ o B 3/4 11<<5 3/4.11.4 TOTAL DOSE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 1 1 5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4 12 1 ~ MONITORING PROGRAM ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 12-1 3/4.12.2 LAND USE CENSUS . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ i ~ ~ ~ . ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM ......, .. ~...... B 3/4 12-1 D. C. COOK - UNIT 2
s INDEX DESIGN FEATURES SECTION PAGE 5.1 'SITE Exclusion Area ........ <<..... ~, ~............ <<. ~.... .. ~ ~ .. 5-1 Low Population Zone ........................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5~1 Site Boundary for Gaseous and Liquid Effluents ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5~1 5.2 CONTAI19ENT Conf iguration ....;, . >>.... . >>..... <<.~ ~ ~ . ~ .. >> ~ ~ . ~ .. ~ ~ ~ ~ ~ ~ 5-1 Design Pressure and .Temperature ........................... 5-1 5.3 REACTOR CORE Fuel Assemblies ........................................... 5-4 Control Rod Assemblies .......,...,........................ 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature ...........,..............'. 5-4 Vo lunule ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ 5~5 5 '5 'METEOROLOGICAL "TOWER LOCATION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 5 5'6 'FUEL STORAGE Criticality - Spent Fuel ..................,............... 5-5 Criticality - New Fuel ...............,......,............. 5-5 Drainage ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 6 Capacity . 1.40.00. 14.0.. 11 ~ 100 11000000 1. 1.0.000000 1000000.. 5-6 5 ~7 COMPONENT CYCLIC OR TRANS IENT LIMIT ~ ~ ~ ~ ~ ~ ~ ~
' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 6 D. 'C. COOK - UNIT 2
INDEX ADf1INISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY....,...............-...................... -.
6. 2 ORGANIZATION f t 0 ffslte ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ ~ ~ . ~ ~ ~ ~ ., ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-1
~
F ac111ty g ~
Staff...............,.;;........................... 6-1 6.3 FACILITY STAFF QUALIFICATIOHS........,....................:, 6-5 6 o4 Dh MTklL TPmINIs>Go ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ e ~ ~ ~ ~ ~ ~
6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW 'COMMITTEE Function......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
Composition...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-6 A1ternates.................... ~ I~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ '0 ~ ~ ~ ~ ~ ~ ~ 6-6 et eet1ng Frequency........... ~.... ~ ~ "~ ~ ~ " ~ ~ "~ ~ ~ ~ " ~ ~ ~ ~ ~ -. 6-6 Q uorum.............,.............................,.........
Responsibilities. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
Authority........ ~ ~, ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-8 R ecords ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-8 6.5.2 NUCLEAR SAFETY AND DESIGN REVIEW 'CONNITTEE unc'0n ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-8 C po 1 1t'mDO>
tl On ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
6 9 Alternates..................,.....,........................ 6a D. C. COOV - UNIT 2 -XVI
ADMINISTRATIVE CONTROLS SECTION PAGE C OnSultantS o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
It
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-9 Meeting Fx'equency . ~.... ~ ~ . ~ ~ .. ~ ~ ~ . ~ . ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-9 Q uo ML ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-10.
Review ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-10 Audits ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-11 AuthOTi 'ty o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6" 12 I ReCOIdS o ~ ~ o ~ ~ o o ~ ~ o o ~ o ~ ~ ~ o ~ o o o ~ ~ o o ~ o o o ~ oo o o o o o ~ ~ o ~ o o'o o o o ~ o o 6-12 6.6 REPORTABLE OCCURRENCE ACTION .......oo...o..o.ooo...oo..ooo os 6-12 6.7 SAFETY LIMIT VIOLATION .. ~ ~ ~ ooo ~ oo ~ ooooooo ~ ~ oo ~ oo,oooo ~ ~ ~ o ~ o 6-13 6 ~8 PROCEDURES ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-1,3 6.9 'EPORTING RE UIREMENTS 6.9. 1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES ........... 6-14 6.9.2 SPECIAL REPORTS 6-22 6o10 RECORD 'RETENTION oooo..oooo ~ oooo..o ~ ~ ~ ~ ooo ~ o ~ ~ ~ ooo ~ ,ohio 6-23 6;11 'RADIATION PROTECTION PROGRAM .......,...................... 6-24 6;12 HIGH RADIATION AREA ...o....... ~ oo ~ o ~ ~ ~ ~ o ~ ~ oooo ~ ~ o ~ oo ~ ~ ~ o ~ ~ 6-24 6.13 ENVIRONMENTAL QUALIFICATION ............................... 6-25 6 14 'PROCESS CONTROL PROGRAM (PCP ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 26 6.15 'OFFSITE 'DOSE CALCULATION MANUAL'ODCM) .........,..., ...... 6-26 6.16 MAJOR CHANGES TO RADIOACTIVE'WASTE TREATMENT 'SYSTEMS ...... 6-27 D. C. COOK UNIT 2 XVII
k t
OEF IH IT IONS UNIOEHTI FIED LEAKAGE r
1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE..
PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE 80UNOARY LEAKAGE shall be leakage {except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.
CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seal s.
QUADRANT POMER TILT RATIO 1;18 QUADRANT POMER TILT RATIO shall be the. ratio, of. the maximum upper excore detector calibrated output to the average of'.the upper excore detector calibrated oQtputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is grea er. Mith one excore detec.or inoperable, the ranaining three detectors shall be used for computing the average.
DOSE EP(UIYALEHT I-13T 1.19 OQSE, EqUIVALEHT I-131 shall be that concentration of I-131 (gCi/gram) which alone would produce the same thyroid dose as the quantity and iso-topic mixture o= I-131, I-132, I-133, I-134, and I-135 ac ually present.
The, thyroid dose conversion factors used or this calculation shall be those listed in Table III of TID-14844, "Calculation of Oistanc Fac ors for Power and Tes~'eactor Sitesor in NRC Regulatory Guide 1.109 Rev. 1, October 1977.
0. C. COOK - UNIT Z
DEFINITIONS PHYSICS TESTS 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 13.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.
E - AVERAGE DISINTEGRATION ENERGY 1.26 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of i,
the sum of the average beta and gamma e'nergies per disnitegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95Ã of the total non-iodine activity in the coolant.
SOURCE CHECK IVI c;c.;EC< '1k h 1i fCI radioactive source.
response when the Channel sensor is exposed to a PROCESS CONTROL PRCGRA)M PCP)
. 8 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analysis, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes will be accomplished in such a way as to assure ccmpliance with 10 CFR 20, 10 CFR 71, Federal and State regulations and other requirements governing the shipment and- disposal of radioactive waste.
D. C. COOK - UNIT 2 1 DEFINITIONS SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of radioactive liquid, resine and sludge wastes from liquid systems into a form that meets shipping and burial site requirements.
OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.30 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents,'n the'alculation of gaseous and liquid effluent monitoring alarm/trip setpoints and the conduct of environmental radiological monitoring program.
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 off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
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 particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulate from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components, PURGE-PURGING 1.33 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 1,34 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.
D. C. COOK UNIT 2 1-7
DEFINITIONS MEMBER(S) OF THE PUBLIC l.35 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the Plant. This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the Plant.
SITE BOUNDARY 1.36 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased or otherwise controlled by the licensee.
UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional and/or recreational purposes'.
C, COOK UNIT 2 1-8
TASLE 1.1 OPERATIONAL iiiQDES REACTIVITY RATED AVERAGE COOLANT vlODE CQNDITICN X.ff THERMAL POWER~ TE;,lP ERATURE
1. POWER OPERATION > 0.99 5w > 350'F.
2. STARTUP > 0.99 < 5% > 350'F
3. HOT STANDBY < 0.99 . > 350 F
4. HOT SHUTDOWN < 0.99 350 F >Tavg
> 200'F
5. COLD SHUTDOWN < 0.99 < 200'F
6. REFUELING~ < 0.95 < 140'F
" Excluding decay heat.
~'Reactor vessel head unbolted or removed and fuel in the vessel.
D. C, COOK - UNIT 2 1-9
TABLE 1.2 FREQUENCY NOTATION NOTATION ~FRE UENCY AT LEAST ONCE PER 12 HOURS AT LEAST ONCE PER 24 HOURS AT LEAST ONCE PER 7 DAYS AT LEAST ONCE PER 31 DAYS AT LEAST ONCE PER 92 DAYS SA AT LEAST ONCE PER 184 DAYS AT LEAST ONCE PER 549 DAYS s/v PRIOR TO EACH REACTOR START-UP COMPLETED PRIOR TO EACH RELEASE N.A. NOT APPLICABLE D. C, COOK UNIT 1-10
C
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I ~ ~ I e
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I
TABLE 3.3-12 RADIOACTIVE LI UID EFFLUENT MONITORING IHSTRUMEHTATION Minimum Channel s Instrument ~0erabl e Action
1. Gross Radioactivity Monitors Providing Automatic Release Termination
a. Liquid Radwaste At times of release 23 Effluent Line
b. Steam Generator At times of release Blowdown Line
c. Steam Generator At times of release Blowdown Treatment Effluent Z. Gross Radioactivity Monitors Not Providing Automatic Release Termination
a. Service Mater System At all times 25 Effluent Line
3. Continuous Composite Sampler Flow Monitor
a. Turbine Building Sump At all times 25 Effluent Line
4. Flow Rate Measurement Devices a." Liquid Radwaste Line (1) At times of release 26
b. Discharge Pipes* (1) At all times NA
c. Steam Generator Blowdown Treatment Effluent (1) At times of release 26
Pump curves and valve settings may be utilized to estimate flow; in such cases, Action Statement Z6 is not applicable.
D. C. COOK - UNIT 2 3/4 3-54
TABLE 3.3-12 (Continued)
TABLE NOTATION Action 23 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may'be resumed for up to 30 days, provided that prior to initiating a release:
1. At least two independent samples are analyzed in accordance with Specification 4. 11.1.1.1:. and;
2. At least two technically qualified members of the Facility Staff independently ver ify the discharge valving. Otherwise, suspend release of radioactive effluents via this pathway.
Action 24 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for gross radioactivjty (beta or gamma) at a limit of detection of at least 10 qci/gram:
l. At least once once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is )0.01 ~ci/gram DOSE EgUIYALENT I-131.
2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is <0.01 pci/gram DOSE EQUIVALENT I-131.
Action 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , grab samples are collected and analyzed for gross radioactivity (bet~ or gamma) at. a lower limit of detection of at least 10 pci/ml.
Action 26 With the number of Channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.
D. C. COOK - UNIT 2 3/4 3-55
TABLE 4.3-8 RAOIOACTIYE LIgUIO EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK: CHECK CALIBRATION TEST I. Gross Beta or Gamma p%'OURCE Radioactivity Monitors providing alarm and automatic iso1ation
~
a. Liquid Radwaste R(3) 0(I)
Effluent Line
6. Steam Generator R(3) a(I)
Blowdown Effluent Line
c. Steam Generator R(3)
Blowdown Treatment Effluent Line
2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm But Not Providing Automatic Isolation
a. Service Mater R(3) a(2)
System Effluent Line
3. Continuous Composite Samplers
a. Turbine Building N/A N/A N/A Sump Effluent Line Flow Rate Monitors
a 0 Liquid Radwaste O(4) N/A Effluent *
b. Steam Generator O(4) N/A N/A N/A Blowdown Treatment Line
During Releases Via This Pathway D. C. COOK - UNIT 2 3/4 3-56
TABLE 4.3-8 (Cont)
TABLE NOTATION The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:
1. Instrument indicates measured levels above the alarm/
trip setpoint.
2. Circuit failure.*
~ 3. Instrument indicates a downscale failure.*
~ 4. Instrument control not set in operating mode.*
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
1. Instrument indicates measured levels above the alarm setpoint.
~* 2. Circuit failure.
3. Instrument indicates a downscale failure. ~
4. Instrument"controls not set in operating mode.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more sources with traceability back to the National Bureau of Standards. These sources shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources 'that have been related to the initial calibration may be used.
(4) CHANNEL CHECK shall, consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic or batch releases are made.
Instrument indicates, but does not provide for automatic isolation.
As equipment becomes operational.
D. C. COOK - UNIT 2 3/4 3-57
Instrumentation Radioactive Gaseous Process and Effluent Monitorin Instrumentation Limitin Condition for 0 eration 3.3.3..10 The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of 3.11.2.1 are not exceeded.
~bbi b111: A 3 1 1 b1 3.3-13.
Action:
a 0 With a radioactive gaseous process or effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of 3.11.2.1 are met. without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, reset, or declare the channel inoperable.
With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.
C. The provisions of Specification 3.0.3, 3.0.4 and 6; 9.1.13 are not applicable.
Surveillance Reouirements 4.3.3.10.1 The setpoints shall be determined in accordance with methodology as described in the OOCM and shall be recorded.
4.3.3.10.2 Each radioactive gaseous process or effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.
This surveillance requirement does not apply to the Waste Gas Holdup System Hydrogen and Oxygen Monitors, as their setpoints are not addressed in the ODCM.
D. C. COOK - UN IT 2 3/4 3-58
TABLE 3.3-13 Radioactive Gaseous Effluent Monitorin instrumentation Minimum Channels Instrument ~0ereb1 e A licabilit Action Maste Gas Holdup System Explosive Gas Monitoring System
a. Hydrogen Monitor (I) 30
b. Oxygen Monitor (2) 29
2. Condenser Evacuation System
a. Noble Gas Activity (I) 28 Monitor
b. Flow Rate Monitor (I) 27
3. Auxiliary Building Ventilation System a.'oble Gas Activity (I) 28 Monitor
b. Iodine Sampler, (I) 32 Cartridge
c. Particulate (I) 32 Sampler Filter
d. Effluent System (I) 27 Flow Rate Measuring Device
e. Sampler Flow Rate 27 Measuring Device
4. Containment Purge System ***
a. Noble Gas Activity (I) **** 31 Monitor
b. Particulate Sampler (I) **** 32 Maste Gas Holdup System
a. .Noble Gas Activity 33 Monitor Providing Alarm and Termination of Gas Decay Tank Releases
6. Gland Seal Exhaust
a. Noble Gas Activity Monitor 28
b. Flow Rate Monitor 27
0. C. COOK - UNIT 2 3/4 3- 59
IABL 3.3-13 (Con.)
At all times orna'c During was e gas holduo system ooeration ('reatment -,or primary system gases).
~ Monitors sample containment atmosphere no- containmen-. pur e. Au termination of. purge on high containment ac. vity.
inuring rel ses via this pa ..':.'ay For purge purooses only, see Tecnnical Specirications 3.3.3.1, 3.-'.=..1, and 3.9.9 for other requirements.
2 For gas decay tank releases only, see 3. for additional requirements.
D. C. COOK - UNIT 2 3/4 3-60
TABLE 3.3-13 (Cont)
TABl E NOTATION Action 27 With the number of channels OPERABLE less than requir d by the Ninimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
Ac.ion 2B With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once=per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
Action 29 With the number of channels OPERABLE one less than required by the Hinimum Channels OPERABLE requirement, operation of this system may continue for up to 14 days. With 2 channels in-operable,,operation of'his system may continue for up to l4 days, provided grab samples are taken and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Ac ion 30 With the number of channels OPERABLE less than required by the i4hnimum Channels OPERABLE requirement, operation of this system may continue for up to 14 days, provided grab samples are taken and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
Action 31 Mith the number of channels OPERABLE less than required by the blinimum Channels OPERABLE requirements, immediately suspend PURGENG of radioac ive effluents via this pathway.
Action 32 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2.
Action 33 With the number of channel's OPERABLE less than required by the Minimum Channels OPERABLE. requirement, the contents of the t nk(s) may be released to the environment for uo to 14 days provided that prior to initiating the release:
a. At least two independent samples of the tank's cont nts are analyz d and,
b. At least two t chnically qualified members of the Faci Iity Staff independently verify the release rate calculations and discharge valve lineups; otherwise, suspend release of radioactive ef luents via this path~ay.
0. C-. COOK - UNIT 2 3/4 3-6j,
TABLE 4.3-9 Radioactive Gaseous Effluent Monitorin Instrumentation Surveillance Requirements Channel Channel Source Functional Instrument Check Check Channel'alibration Test:.
Maste Gas Holdup System Explosive Gas Monitoring System
a. Hydrogen Q(3)
'onitor
b. Oxygen Q(4)
Monitor
c. Oxygen NA Q(4)
Monitor(Alt.)
Condenser Evacuation System
a. Noble Gas Activity 0*
R(2) Q(1)
Monitor
b. System Effluent 0*
NA Flow Rate Auxiliary Building Ventilation System
a. Noble Gas Activity 0* R(2) Q(1)
Monitor
b. Iodine Sampler M* NA NA NA C. Particulate Sampler M* NA NA NA
d. System Effluent 0* NA R Q Flow Rate Measure-ment Device I
e. Sampler Flow Rate D* 'NA Measurement Device Containment Purge System
a. Noble Gas Activity 0~ R(2) Q(5)
Monitor
b. Particulate Sampler M~ NA NA NA Maste Gas Holdup System
a. Noble Gas Activity P** R(2) Q(5)
, Monitor Providing Alarm 8 Termination of Gas Decay Tank Releases D. C. COOK - UNIT 2 3/4 3 62
TASLr < 3 a ro~t
o. Glanc Seal =.~hausz
a. io'."le Gas Act-'v ty
'loni tor
b. System -"ffluent NA Flow Rate At all times Ouring Release Yia This Pathway
"** Ouring Maste Gas Holdup System Operation (Treatment for Primary System Offgases)
D. C. COOK - UNIT 2 3/4 3-63
TABLE 4.3-9 (Cont)
TABLE NOTATION The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions ex i sts:
Instrument indicates measured levels above the alarm setpoint.
Circuit failure.
Instrument indicates a downscale failure.
Instrument controls not set in operate mode.
2) The initial CHANNEL CALIBRATION shall be performed using one or more sources with traceability back to the National Bureau of Standards. These sources shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial. calibration may be used.
3) 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.
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1. One volume percent oxygen, balance nitrogen, and
2. Four volume percent oxygen, balance nitrogen.
.s) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm'nnunciation occurs if. any of the following conditions exists:
I. Instrument indicates measured levels above the alarm/trip setpoint.
2. Circuit failure.*
~3. Instrument indicates a downscale failure.*
~4. Instrument controls not set in operate mode.*
Instrument indicates, but does not provide automatic isolation.
As equipment becomes operational.
D. C. COOK - UNIT 2 3/4 3-64
INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4.1 At least one turbine overspeed protection system shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
a. With one stop valve or one control valve per high pressure turbine steam lead inoperable or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam lead inoperable, operation may continue for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the inoperable valve(s) is restored to OPERABLE status or 'at least one valve in the affected steam lead is closed; otherwise, isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b. With the above required turbine overspeed protection system otherwise inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either restore the system to OPERABLE status or isolate the turbine from the steam supply.
SURVEILLANCE RE UIREMENTS 4.3.4.1.1 The provisions of Specification 4.0.4 are not applicable.
4.3.4..1.2 The above required turbine overspeed protection system shall be demonstrated OPERABLE:
a. At least once per 7 days by cycling each of the following valves through at least one complete cycle from the running position;
1. Four high pressure turbine stop valves.
2. Four high pressure turbine control valves.
3. Six low pressure turbine reheat stop valves.
4. Six low pressure turbine reheat intercept valves.
D. C. COOK - UNIT 2 3/4 3-65
INSTRUMENTATION LIMITING CONOITION FOR OPERATION
b. At least once per 31 days by direct observation of the move-ment of each of the above valves through one complete cycle from the running position.
c~ At least once per 18 months by performance of a CHANNEL CALIBRATION on the turbine overspeed protection systems.
d. At least once per 40 months by disassembling at least, one of each of the above valves and performing a visual and surface inspection of valve seats, disks and stems and verifying no unacceptable flaws or corrosion.
0. C. COOK - UNIT 2 3/4 3-66
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 Li uid Effluents Concentration Limi tin Condition for 0 eration 3.11.1.1 The concentration of radioactive material released at any time from the site to unrestricted areas (see Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix 8, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases.
For dissolged or entrained noble gases, the concentration shall be limited to 2 X 10 pci/ml total activity.
A Action:
>lith the concentration of radioactive material released from the site ex-ceeding the above limits, without delay restore the concentration to wTthin the above limits.
SURVEILLANCE REOUIREHENTS
4. ll. 1. 1. 1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4. 11-1.
4.11. 1. 1.2 The result of radioactive analysis shall be used in accordance with the methods of the ODCH to assure that all concentrations at the point of release are maintained within the limits of Specification 3. 11. 1. 1.
D. C. COOK - UNIT 2 3/4 11-1
TABLE 4. 11-1 Radioactive Liquid Waste Sampling and Analysis Program Lower Limit a Minimum Type Of of Detection Liquid Release Sampling Analysis Activity (LLD)
T e Fre uenc Fre uenc Anal sis >ici/ml A. Batch Waste Release Tanks Principal 5X10 Gamma Emitters Each Batch Each Batch I-131 lxlo Di ssol ved 1X10 and Entrain-ed Gases (Gamma
. One Batch/M Emitters) p M H-3 1xlo Batch Composite Gross Alpha x 1 0-
'ach 1
p b Sr-89,Sr-90 5X10
'ach Batch Compositi= re- 1X10 B. Plant Continu'ous Daily W Principal 5X10 Releases d Gamma b Emitters Composite I-131 1 xl 0 Dissolved 1X10 Grab Sample and Entrain-ed Gases (Gamma Emitters)
Daily M b
H-3 1X10 Com osite 'ross Al ha 1X10 Daily Q b Composite Sr-89, Sr-90 5X10 Fe-55 1X10-6 D. C. COOK - UNI'T 2 3/4 11-2
TABLE 4.11-1 (Cont)
TABLE NOTATION
a. The lower limit of detection (LLD) is'efined in Table Notation
a. of Table 4.12-1 of Specification 4.12.1.1.
b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is re-presentative of the liquids released.
c. A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analysis, each batch shall be isolated and re-circulated to ensure thorough mixing.
d. A continuous release. is the discharge of liquid waste of a non-discrete volume; e.g. from a volume of system that has an input flow during the continuous release.
e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Nn-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, together with the above nuclides, shall also be identified and reported.
D. C. COOK - UNIT 2 3/4 11-3
I I I I I I
~ ~ e ~ I ~ ~ ~ ~ ~
I ~ I ~ ~ ~ ~ ~ ~ 0
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II
A Radioactive Effluents Li uid Waste Treatment Limitin Condition For 0 eration 3.11.1.3 The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the pro-jected doses due to the liquid. effluent from the site (see Figure 5.,1-3) when averaged over 31 days', would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.
baal i: A Action:
a. Mith radioactive liquid waste being discharged'ithout treatment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:
Identification of the inoperable.
J'.
equipment or subsystems and the reason for inoperabi1 i ty,
2. Action(s) taken to restore the inoperable equipment to operable status, and
3. Summary description of action(s) taken to prevent recurrence.
b. The provisions of Specification 3.0.3, 3'.0,.4 and'.9;1'.'13:.a'e not..
applicable.
Surveillance Re uirements 4.11.1.3 Doses due to liqiid releases to UNRESTRICTED AREAS shal.l be pro-jected at least once per 31 days, in accordance with the ODCl1, whenever liquid releases are being made without being orocessed by the liquid rad-waste treatment system.
D. C. COOK - UNIT 2 3/4 Sl-5
Radioactive Effluents Li uid Holdu Tanks*
Limitin Condition For 0 eration
3. 11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited to less .than or equal to lg curies, excluding tritium and dissolved or entrained noble gases.
a. Outside temporary tanks.
A Action:
a. Mith the quantity of radioactive material in any of the above listed tanks exceeding the above limi t, without delay suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to wi thin the limit.
b. The provisions of Specification 3.0.3, 3. 0.4 and 6; 9.,1.'13 are, not
'pplicable.
Surveillance Re uirements
4. 11. 1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by .
analyzing a representative sample of the tank.'s contents at least once per 7 days when radioactive materials are being added to the tank.
Tanks included in this Specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and sur-rounding area drains connected to the liquid radwaste treatment system.
D. C. COOK - UNIT 2 3/4 ll-.o
Radioactive Effluents 3/4.11.2 Gaseous Effluents Dose Rate Limitin Condition For 0 eration 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site (See Figure 5.1-3) shall be limited to the following:
a. For noble gases: < 500 mrem/yr to the total body and < 3000 mrem/yr to the skin, and
b. For all radioiodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than 8 days: < 1500 mrem/yr to any organ.
A Action:
With the dose rate(s) exceeding the above limits, without delay decrease the release rate to within the above limit(s).
Surveillance Re uirements 4.11.2.1.1 The dose rate due to'oble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCN.
4.11.2.1.2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCN by ob-taining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.
~
D. C. COOK
~ ~
- UNIT 2 3/4 11-7
~ ~
~ I ~
I
~ ~
~ ~ I I
I
~ I
~ ~
~ ~ 'I ll ~
~ I~
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TABLE 4.11-2 (cont)
TABLE NOTATION
a. The lower limit of detection (LLD) is defined in Table Notation a.
of Table 4.12-1 of Specification 4.12.1.1.
occurrence
b. Analyses shall also be perfor'med following any operational which has altered the mixture of radionuclides as indicated by RCS analysis. (ie., start-up.)
c. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing. Analys'es shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 .days following each shutdown, startup or similar operational occurrence which lead to significant increases or decreases in radioiodine in the Reactor Coolant System. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor of 10.
d. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specification 3. 11.2.1, 3.11.2.2, and 3.11.2.3.
e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133M, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and reported.
D. C. COOK - UNIT 2 3/4 11-9
RAD IOACTIVE EFFLUENTS DOSE, NOBLE GASES LIMITING CONDITION FOR OPERATION
3. 11.2.2 The air dose in UNRESTRICTED AREAS due to noble gases released in gaseous effluents shall be limited to the following:
a. During any calendar quarter,, to < 5 mrad for. gamma radiation and < 10 mrad for beta radiat,ion;
b. During any calendar year, to < 10 mrad for gamma radiation and
< 20 mrad for beta radiation.
APPLICABILITY: At al 1 times.
ACTION:
a. Mith the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare.
and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report whi'ch identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent re-
'leases will be within the above limits.
b. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for the total time period sha be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least once every 31 days.
D. C. COOK - UNIT 2 '/4 11-10
RADIOACTIVE EFFLUENTS DOSE RADIOIODINES RADIOACTIVE MATERIAL IN PARTICUL'ATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to A >EMBER OF THE PUBLIC from radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days in gaseous effluents released to unrestricted areas shall be limited to the following:
a. During any calendar quarter to ~ 7.5 mrem to any organ;
b. During any calendar year to ~ 15 mrem to any organ.
APPLICABILITY: At all times'CZ ION:
aO With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Spec'al Report which identifies the cause(s) for exceeding the limit and defines the corrective actions taken to reduce the releases and the proposed corrective action to be taken to assure that subsequent releases will be within the above limits.
b, The provisions of Specification 3.0.3, 3.0.4, and 6.9.1.13 are not applicable.
SURVEILLANCE 'RE UIREMENTS 4,11.2.3 'OSE CALCULATIONS Cumulative dose contributions for the total time period shall be determined in accordance with the ODCM at least once every 31 days.
D. C. COOK - UNIT 2 3/4 11-11
Radioactive Effluents Gaseous Radwaste Treatment Limitin Condition For Ooeration 3.11.2.4 The gaseous radwaste treatment system and the ventilation exhaust treatment system shall be used to reduce the radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent rel,eases to unrestricted areas (See Figure 5.1.3) when averaged over 31 days, would exceed 0.2 mrad for ganja radiation and 0.4 mrad for beta radiation. The ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1-3) when averaged over 31 days would exceed 0.3 mrem to any organ.
11111.': 1 Actian:
a. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission .within 30 days, pursuant to Specification 6.9.2, a Special Report which in-cludes the following information:
1. Identification of the inooerable equipment or subsystems and the reason for. inoperabil.ity-.
2. Action(s) taken to restore the inoperable equipment to operable status.
b. The provisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicabIe.
Surveillance Reouirements 4.11.2.4 Doses due to gaseous releases to UNRESTRICTED AREAS shall be pro'ecte at leas once per 31 days in accordance with the ODC<, whenever the gaseous was.e treatment systen or ventilation exhaust treatment system is not operational.
D. C. COOK - UNIT 2 3/4 j'-12
RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION
3. 11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to < 2% by volume if the hydrogen in the system is > 4% by volume.
APPLICABILITY: At all times.
ACTION:
a. With the concentration of oxygen in the waste gas holdup system
> 2% by volume but < 4% by volume and containing > 4/ hydrogen, restore the concentration of oxygen to < 2% or reduce the hydrogen concentration to < 4% within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .
b. With the concentration of oxygen in the waste gas holdup system or tank > 4% by volume and > 4% hydrogen by volume without delay suspend all additions of <<'aste gases to the system or tank and reduce the.
concentration of oxygen to < 2% or the concentration of hydrogen to
< 4% witnin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months in the system or tank..
c. The provisions of Specification'.O.3, 3.0.4 and 6.9,1.13 are not applicable.
SURVEILLANCE RE UIREMENTS
4. 11.2.5 The concentration of oxygen in the waste gas holdup system shall be determined to within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.10.
D. C. COOK - UNIT 2 3/4 11-13
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SOLID RADIOACTIVE WASTE SURVEILLANCE REQUIREMENTS 4.11.3.1 The solid radwaste system shall be demonstrated operable at least once per 92 days by:
a. Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or
b. Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.
.11.3.2 THE PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION f at least one representative test specimen from at least every tenth atch of each type of wet radioactive waste (e.g. filter sludges, spent esins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions).
a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with .the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.
b; .If'he" initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CQNTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste unti:1 at least 3 consecutive ini till test shall specimens demonstrate modified SOLIDIFICATION. The PROCESS CONTROL PROGRAM be as required, as provided in Specification 6.14 to assure SOLIDIFICATION of subsequent batches of waste.
D. C. COOK - UNIT 2 3/4 11-16
RADIOACTIVE EFFLUEHTS 3/4 11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to c 25 mrem to the total body or any organ (except the thyroid, which is lim7ted to < 75 mrem) over a period of 12 consecutive months.
APPLICABILITY: At all times.
ACTION:
a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3. 11.1.2.a, 3. 11.1.2.b, 3. 11.2.2.a, 3.11.2.2.b, 3. 11.2.3.a, or
3. 11.2.3.b, in lieu of any other report required by Specification 6.9'. 1, prepare and submit a Special Report to the Director," Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D. C. 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3. 11.4. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 12 consecutive month period that includes the release(s) covered by this report. If the estimated dose(s) exceeds the limits of Specification 3.11.4, and if the release condi'tion resulting in violafion of, 40 CFR 190 has not already been corrected, the Special Report shall include .a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified infor-mation of 5 190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as addressed in other sections of this Technical Specification.
b. The provisions of Specification 3.0.3, 3. 0.4 and 6. 9. l. 13 are not applicable.
SURYEILLANCE RE UIREMENTS 4.11.4 g ~h Dose Calculations 4.11. 1.2, 4. 1 1.2. 2, and 4, 1 Cumulative dose contributions from g
l. 2.3 g gg and wi th the ODCH.
ggg liquid hagi and D. C. COOK - UNIT 2 3/4 11-17
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RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION CONTINUED)
c. With milk or fresh leafy vegetable samples unavailable from any of the sample locations required by Table 3.12-1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause of the unavaila-bility of samples and identifies locations for obtaining replace-ment samples. The locations from which samples were unavailable may then be deleted from Table 3. 12-1 provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations, available.
if
d. The provisions of Specification 3.0.3, 3.0.4 and 6.9,l.l3 are not applicable.
SURVEILLANCE RE UIREMENTS
4. 12. 1 The radiological environmental monitorincf samples sha)l be collected pursuant to Table 3. 12-1 from the locations given in the table and figures in the ODCM and shall be analyzed pursuant to the requirements of Tables 3.12-1 and 4.12-1.
D. C. COOK - UNIT 2 3/4 12-2
1'ABLE
~ ~
3.12-1'ADIO'G:CAL E I'lIROHMENTAL MONITORING PROGRAM Exposure . a"hway Sampl ng and Type of Fl'equency and/or amc s "am"le L cations Col',ac=ion Freouencv oT Alla l vs~i s
l. Airborne
a. Radioiodine AI-A6 (Site) Continuous operation Radioiodine canis and Particulates of sampler with Analyze New BuT'alo, Sample Collection Meekly for I-131 South Bend as required by
, Oowagiac, and Oust Loading But At Particulate sampler:
Coloma are Least once Per 7 Days Gross Beta Rad-Background ioactivity followinc Filter Changea , cor.
posi te (by loca-tion) for gamma isotopic quarterly
2. Direct Radiation a) TI-T9.(Site) At least once per Gamrr~ Dose. At 92 days Leas" Once Per New Buffalo 92 days South Bend Oowagiac Coloma
.. c) 10 TLD Monitor Locations in the Five Mile Radius
3. Materborne
a. Surface LI , LZ , L3 Compos> te Sampl e Gamma Isotopic Over One-Month Period Analysis monthly .
Ccmposite For tritium analysis quarter I y.
b. Ground M14 7 quarterly Gamma isotcoic and Tritium analysis guarteri y .
c. Drinking St. Joseph Ccmposit~ Sample Gross Beta and Lake Township Collected over a Gamma Isotopic New Buffalo Period of <3I days Analysis of each Ccmposfte" Sample ccmposite sample.
Over a 2 week Tritium Analysis Period if I-131 of ccmccsite Analysis is Performed. quarterly.
I -131 analysis on each ccmcosi-e when the dose calcuia ed Ccmcosite samples shall be collected by collecting an aiicuot or the consumption a t irtervals not exceeding.24 hcurs.
jer year.
D. C. COOK - UNIT 2 3/4 12-3
TABLE 3.12-1 (Cont)
d. Sediment from L2, L3 2/year Gamma Isotopic Shoreline Analyses Semi-Annual 1y .
4. In9estion a.Mi1k Stevensville At least once per Gamma Isotopic Bridgman 15 days when animals and I-131 Analysis Gal i en are on Pasture. At of Each Sample.
Dowagiac Least Once Per 31 South Bend . Days at Other Times .
b. Fi sh Plant Site 2/year Gamma Isotopic Off-Si te Analysis on Edible Portion.
c. Food Products Plant Site At time of Harvest Gamma Isotopic Off-Site (approx. One Sample of Each Analysis on Edible
.20 mi) of the Following Portion.
Classes of Food Products
1. Grapes Plant Site At time of Harvest Gamma Isoto One sample of Broad Analysis Leaf Yegetation aParticulate sample filters should be analyzed for gross beta 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 times the yearly mean of control samples for any medium, gamma isotopic analysis should be performed on the .individual samples.
D. C. COOK - UNIT 2 3/4 12-4
ABLe 3 12 2 Reporting Levels For Radioactivit'J Concen7rations In Environmental Samples Mater Airborne ParticgEate Fi h Milk Food Prod Analjsis (pci/1) or Gases (pci/m 1 (pcigg,wet) (pci/1) (pci/kg,wet H-3 2 X 10 Nn-54 1 X10 3 X 10 Fe-59 4 X 10 1X10 3
Co-58 1X10 3X10 4
CO-60 3X10 2 1 x 10 Zn-65 3X10 2 X 10 2
Zr-Nb-95 4 X 10 I-131 0.9 1X10 3
Cs-134 30 10 1X10 60 . 1X10 3
. Cs-137 50 20 2X10 70 2X10 2
Ba-La-140 2 X 10 3X10
0. C. COOK - UNIT 2 3/4 12-5
TABLE 4.12-1 Maximum Values For The Lower Limits of Oetection (LLO) '
Airborne Particulate Analysis Mater or Fish Milk Food Products Sediments (pc i/1) Gas 3
(pci/kg) (pci/1) (pci/kg,wet) (pci/kg,dry (pci/m ) wet Gr oss Beta 1X10 H-3 2000 Mn-54 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-95 30 Nb-95
-2 I-T31 ?X10 60
-2 Cs-134 15 6 X 10 130 60 150 Cs-13? 6 X 10 150 18 60 180 Ba-140 La-140 15 O. C. COOK - UNIT 2 3/4 12-6
TABLE 4.12-1 (Cont)
TABLE NOTATION
a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95K probability with 5/ probability of fal sely concluding that a blank observation represents a "real" signaI.
For a particular measurement system (which may include radiochemical separation):
4 66 sb LLD =
E
' 2.22 ~
Y ~
exp (-Rat) where LLD is the "a priori" lower limit of detection as defined above (as pci per unit mass or volume),
s is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency(as counts per transformation),
Y is the sample size (in units of mass or volume),
2.22 is the number of transformation per minute per picocurie, Y is the fractional radiochemical yield (when applicable),
X is the radioactive decay constant for the particular radionuclide and
~t is the elapsed time between sample collection (or end of the sample collection period) and time of counting.
The volume of Sb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appro-priate) rather than on an unverified theoretically predicted variance.
In calculating the LI D for a radionuclide determined by gamma-ray spec-trometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., potassium - 40 in milk samples).
0. C. COOK - UNIT 2 3/4 12-?
Table 4. 12-1 (Cont)
Table Notation Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of inter-ferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contribution factors will be identified and described in the Annual Radiological Environmental Operating Report.
b. LLD for drinking water.
c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identified and reported.
D. C. COOK - UNIT 2 3/4 12-8
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPFRATION 3.12.2 A land use census shall be conducted and shall ide'ntify the loca-tion of the nearest milk animal, the 'nearest residence and the nearest garden* of greater than 500 square feet producing fresh leafy vegetables in each of the . 9 land covering meteroloqical sectors within a distance of five miles.
APPLICABILITY: At al 1 times.
ACTION:
a. With a land use census identifying a location(s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, pre-pare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the ne>v location(s).
f
b. With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exoosure pathway) 20 percent areater than at a location from which samples are currently being obtained in accordance with Specification 3. l2.1, prepare and submit to the Comm1ssion within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new location. The new location shall be added,to the radiological environmental monitoring program within 30 days, if possible.
The sampling location having the lowest calculated dose or aose commitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the. year in which this land use census was conducted.
c. The orovisions of Specification 3.0.3, 3.0.4 and 6.9.1.13 are not applicab'le.
SURVEILLANCE RE UIREMENTS 4.12.2. The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1, by door-to-door survey, aerial survey, or by consulting local agriculture authorities.
f f the direction sector with the highest D/g in lieu of the garden census.
0. C. COOK - UNIT 2 3/4 12-9
Radiolo ical Environmental f1onitorin 3 4 12.3 Interlaborator Comparison Pro ram Limiting Condition For Operation 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission.
Aoolicabilit : At all times.
Action:
a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Repor t.
b. The provisions of Specification 3.0.3, 3.0,4 and 6.9.1:13 are not applicable.
Surveillance Re uirements r
4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program and in accordance with the ODCM (or participants in the EPA crosscheck program shall provide the EPA program code designation for the unit) shall be included in the Annual Radiological Environmental Operating Report.
D. C. COOK - UNIT 2 3/4 12-10
3/4 . 3 INSTRUMENTATION BASES 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.
In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. Use of containment temperature monitoring is allowed once per hour if containment fire detection in inoperable, 3/4.3.3.9 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods 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 specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.
3/4.3;3.10 RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm/trip setpoints 'for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.
3/4.3.4 'TURBINE OVERSPEED 'PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.
~
Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.'.
C. COOK - UNIT 2 B 3/4 3-3
3/4.11 RADIOACTIVE EFFLUEHTS 8ASES 3/4.H.1 LIQUID EFFLUENTS 3/4.11;1.1 CONCENTRATION. This specification is provided to ensure that the concentration of radioactive materials rel eased in liquid waste eifluents frcm the site to unrestricted areas will be less than the concentration levels speci-.
fied in 10 CFR Part 20, Appendix 8, Table II. This limitation provides additional assurance that the levels of'adioactive materials in bodies of water outside the site will not result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for noble oases is based upon the assumption that Xe-135 is the controlling radioisotope and its NPC in air (submersion) was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protec-tion (ICRP) Publication 2.
3/4.11.1.2 OQSE. This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Qperaton implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will. be kept "as lowas is ..--
reasonably achievable." Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 4G CFR 1'4l.
The dose calculations in the QOQI implemen the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be show'n by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be subs .antial ly underestimated. The equations specified in the QOCN for calculating the doses due .o the actual release rates of radioactive materials in liouid ef,luents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Ooses to Man from Routine Releases of Reactor Effluents for the Purpose of Evalua ing Compliance with 10 CFR Part 50, Appendix I,"
Revision 1, Cc.ober '1977, and Regulatory Guide 1.113, "Estima ing Aquatic Oispe. sion of E,fluents fr m Accidental and Routine Reactor, Releases for the Purpose of Imp'lementing Appendix I," April 1977. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guide 1.109 and 1.113.
D. C. COOK - UNIT 2 8 3/4 11-1
RAD IQACl:I'lE EFFUJEt~rTS BASES This specification applies to the release of liquid effluents from each reactor at the site. The liquid effluents from the shared sys.em are proportioned among the units sharing the system.
3/4.11.1.3 LIQUID MASTE TREATiilENT. The operability of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system be used when speci-fied provides assurance that the releases of radioactive materials in liquid effluents wi 11 be kept "as low as is reasonably achievable." This speci-fication implements the requirements of 10 CFR Part 50.36a, General Design Criteria Section 11. 1 of the Final Safety Analysis Report for the Donald C.
Cook Nuclear Plant, and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section,II.A of Appendix I, 10 CFR Part 50, for liquid effluents.
3/4.11.1.4 LI UID HOLDUP TANKS Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the, event of an uncontrolled release of the tanks'ontents; the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestric.ed area.
3/4. 11.2. GASEOUS EFFLUDi TS 3/4.11.2.1 DOSE RATE. This specification is provided to ensure that the dose rate any any time at the SITE BOUNDARY from gaseous effluents frey all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix 8, Table II. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, to annual average concentrations exceeding the limits specified in Appendix 8, Table II of 10 CFR Part 20 (10 CFR Part 20.106 (b) ).
For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the aunospheric dif usion factor above that for the site boundary.
The specified release rate limits restrict, at all times, the corre-sponding gamma and beta dose rates above background to an individual at'or beyond the site boundary to < (500) mrna'/year to the total body or to <(3000) mrem/year to the skin. Tnese release rate limits also restrict, at all times, the corresponding thyroid dose rate above back-ground to an infant via the cow-milk-infant pathway to < 1500 mrem/year for the nearest cow to the plant.
This specification applies to the release of gaseous effluents from all reac.ors at the site. The aaseous effluents from the shared system are proportioned among the units sharing tha. system.
0. C. COOK - UNIT 2 B 3/4 11-2
RADIOACTIVE EFFLUENTS BASES 3/4.11.2.2 DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections II.B, III.A and IY.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements pi.ovide the required operating flexibility and at the same time implement the guides set for th in section IV.A of Appendix I to assure that the releases of radio-active material in gaseous effluents will be kept "as low as is reasonaole achievable". The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by calculational procedures based on models and data such that the actuaI exposure of an individual through'he appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be 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 Atmos-pheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Hater-Cooled Reactors," Revision 1, July 1.977... The ODCM '
equations-provided for'etermining the air doses at the site boundary will be based upon the historical average atmospherical conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.
3/4.11.2.3 DOSE, RADIOIODINES, RADIOACTIVE MATERIAL IN PARTICULATE FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES. This specification is provided to implement 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 Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods approved by the NRC 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, "Calculation af 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 ef D. C. COOK - UNIT 2 8 3/4 11-3
RADIOACTIVE EFFLUENTS BASES aseous Effluents in Routine Releases from Light-Mater-Cooled Reactors,"
Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.
The release rate specifications for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependant on the existing radionuclide pathways to man, in the unrestricted area.
he pathways which are examined in the development of these calculations re: '1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat pro-ducing 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 MASTE TREATMENT The operability of the gaseous radwaste treatment system and the ventila-tion exhaust treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion Section ll. 1 of the Final Safety Analysis Report for The Donald C. Cook Nuclear Plant and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
3/4 11. 2. 5 Explosive Gas t4ixture t
This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treat-ment system is maintained below the flammability limits of hydrogen and oxygen mixtures. t5aintaining the: concentration of hydrogen or oxygen below their flammability limits provides that the releases of radioactive materials will be controlled ir, conformance with the requirements of the General Design. Criterion specified in Section 11. I of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.
3/4 11.2.6 Gas Stora e Tanks Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an in-dividual at the nearest site boundary will not exceed 0.5 rem. This is consistant with Standard Review Plan 15.7.1. "Maste Gas System Failure."
D. C. COOK - UNIT 2 8 3/4 11-4
RADIOACTIVE EF FLUENTS BASES 3/4.11. 3 SOLID RADIOACTIVE WASTE The operability of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to shipment offsite, This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion specified in Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limit-ed to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents, mixing and curing time.
3/4.11.4 TOTAL DOSE The specification is provided to meet the dose limitations of 40 CFR 190. The speci ication requires the preparation and submittal of a Special Report whenever the calculated doses from plant r'adioactive effluents exceed twice the design objective doses of Appendix I. For sites reactors, it is highly unlikely that the resultant dose to acontaining up to 4 member 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 describe a course of action which should result in the limitations will member of the public for 12 consecutive months to of dose to a within For the purposes of the Special Report, it may be assumed that the the 40 CFR 190 limits.
ment to the member of the public from other .uranium fuel dose commit-cycle sources is negligible, with the exception that dose contrib~tions from other nuclear fuel cycle facilities at, the same site or within a radius of '5 miles be con-sidered. If the dose to any member of the public is estimated tomust exceed requirements of 40 CFR 190, the Special Report with a request for a variance the (provided the release conditions resulting in violation of already been corrected), in accordance with the provision of40 40CFRCFR190 have not is considered to be a timely request and fulfills the requirements of190.11 40 CFR 190 until NRC staff action is completed. An individual is not considered a member of the public during any period in which out any operation which is part of the nuclear fuel he/she is engaged in carrying cycle.'.
C. GOOK - UNIT 2 B 3/4 11-5
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological monitoring program required by this specification provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program,thet eby supplements the radiological effluent monitoring program by verifying that the measurable concentration of radioactive materials'nd levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.
The detection capabilities required by Table 4.12-1'are the state-of-the art for routine environmental measurements in industrial laboratories.
It should be recognized that the LLD is defined an an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.
Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interferring nuclides, or other uncontrollable circumstances may render these LLDs unachievable.
In such ca'ses, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.
3 4.18.2 LAND USE CENSUS This specification i's provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. This census satisfies the requirements of Section I~l.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (25 kg/year) of leafy vegetables assumed in Regulatory Guide 1,109 for consumption by a child.
To determine this minimum garden size, the following assumptions were used, 1) that 20/ of the garden was used for growing broad leaf vegetation (i.e. simi liar to lettuce and cabbage), and 2) a vegetation field of 2 kg/square meter.
3 4.12.3 INTERLABORATORY CGMPAPISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmentaI monitoring in order to demonstrate that the results are reasonably valid.
0. C. COOK - UNIT 2 8 3/4 12-1
5. 0 DESIGN FEATURES
5. 1 SITE Exclusion Area 5.1.1 The exclusion area shall be shown in Figure 5.1-1.
Low Po ulation Zone
5. 1.2. The low population zone shall be shown in Figure 5. 1-2.
I Site Boundar For Gaseous and Li uid Effluents I
5.1.3 The site boundary for gaseous and liquid effluents shall be shown in Figure 5.1-3.
5. 2 CONTAINMENT CONF I GURAT ION 5.2. 1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:
a. Nominal inside diameter = 115 feet.
b. Nominal inside height = 160 feet.
c. Minimum thickness of concrete walls = 3'".
d. Minimum thickness of concrete roof = 2'".
e. Minimum thickness of concrete floor pad = 10 feet.
f. Nominal thickness of steel liner = 3/8 inches.
g. Net free volume = 1.N x 10 cubic feet.
QESIGN PRESSURE ANO TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained in accordance with the original design provisions contained in Section 5.2.2 of the FSAR.
D. C. COOK - UNIT 2 5-1
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I J
6.0 ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Protection Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1,8, September 1975, and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in Plant design, and response and analysis of the Plant for transients and accidents.
6.3.2 Until the newly appointed Operations Superintendent obtains a Senior Reactor Operator's License, all of his licensed functions will be performed by a full time assistant who holds a current Senior Reactor Operator's License.
6.4 TRAINING 6,4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5 .5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.4.2 A training program for the Fire Brigade shall be maintained under the direction of-the Plant Manager and shall meet or exceed the requirements of Section 27 of the NFPA Code>>1976.
D. C. COOK - UNIT 2
ADMINISTRATIVE CONTROLS 6.5 REVIEM AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIBC COMMITTEE (PNSRC)
FUNCTION 6.5. 1.1 The PNSRC shall function to advise the Plant Manager on all matters reIated to nuclear safety.
COMPOSITION
6. 5.1. 2 The PNSRC shall be composed of the:
Chairman Plant Manager or designated alternate Member: Assistant Plant Managers Member: Operations Superintendent Member: Technical Superintendent Member: Maintenance Superintendent Member: Control and Instrument Supervisor Member: Nuclear/Computer Engineering Supervisor Member: Plant Chemical Supervisor Member: Performance Supervising Engineer Member: Plant Radiation Protection Supervisor Member: Shift Supervisor Member: Environmental Coordinator ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PNSRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PHSRC activities at any one time.
MEETING FRE UENCY 6.5.1.4 The PNSRC shall meet at least once per calendar month and as convened by the PNSRC Chairman or his designated alternate.
gUQRUM 6.5.1.5 A quorum of the PNSRC shall consist of the Chairman or his designated alternate and four members including alternates..
0. C, COOK - UNIT 2 6-6
ADMINISTRATIVE CONTROLS RESPONSIBILITIES
\
6.5.1.6 The PNSRC shall be responsible for:
a. Review of (1) all procedures required by Specification 6.8 and changes thereto, (2) any other proposed procedures or changes thereto as determined by the Plant Manager 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 proposed changes or modifications to Plant systems or equipment that affect nuclear safety,
e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairman of the NSDRC.
f. Review of events r'equiring 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the Commission,
g. Review of facility operation to detect potential nuclear safety hazards.
h, Performance of special reviews, investigations or analysis and reports thereon as requested by the Chairman of the NSDRC.
i. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.
Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.
k. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering-evaluation and recommendations to prevent recurrence to the NSDRC.
1. Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems.
D. C. COOK UNIT 2 6-7
ADMINISTRATIVE CONTROLS AUTHORITY
6. 5.1. 7 T he PNSRC shall;
a. Recommend to the Plant Manager written approval or 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 unreviewed safety question.
c~ Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the NSDRC of disagreement between the PNSRC and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The PNSRC shall maintain written minutes of each meeting and copies shall be provided to the Chairman of the NSDRC.
6.5.2 NUCLEAR SAFETY AND DESIGN REVIEW COMMITTEE NSDRC FUNCTION 6.5.2.1 The NSDRC shall function to provide independent review and audit of designated activities in the ardas of:
a. nuclear 'power plant operations
b. nuclear engineering C. chemistry and radiochemistry
d. metallurgy
e. instrumentation and control radiological safety
g. mechanical and electrical engineering quality assurance practices D. C. COOK - UNIT 2 6-8
AOMINI STRATI'/E CONTROLS COMPOSiTION 6.5.2.2 The NSDRC shall be ccmposed o7 the:
Chairman: Assistant Vice President, Nuclear Engineex'ing Member: Vice Chairman, Engineering and Construction Member: President and Chief Operating Officer of Z&HECo Member: Executive Vice President, Construction and New York Engineering Member: Vice President, Mechanical Engineering Member: 'ice President, Electrical Engineering Member: Vice President, Engineering Administration Member: Assistant Vice President, Design Division Member: Assistant Vice President, Environmental Engineering Division Member: Plant ~i ger, D. C. Cook Plant Member: i~ger, Nuclear Safety and Licensing Section Alternate: Assistant Chief Mechanical Engineer Alternate: Assistant Plant Manager, D. C. Cook Plant Alternate: Executive Assistant to the President of E&WCo Alternate: Assistant Division Manager, Nuclear Engineexing ALTERNATES 6.5.2.3 All alternate members shall be aopointed in writino by the NSORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NSDRC activities at any one time.
CONSULTANTS
.6.5.2.4 Consultants shall be utilized as determined by the .'!SDRC Chairman to provide evoer: advice to the NSDRC.
MEE ING FREOUENCY 6.5.2.5 The NSDRC shall meet at least once per calendar quar;er Curing the initial year or -acility operation =ollowing iuei loading and at least once per six months therea-..er.
D. C. COOK - U" IT 2 5 Q
I AGHINISTRAT /E CONTROLS UD ITS 6.5.2.8 Audits of facility aciivities shall be performed under ihe cognizance of the NSDRC. These. audi is shall encompass:
a. The conformance of facility operation to provisions contained within the Technical Specification and applicable license conditions at leas i once per .12 months.
b. The performance, training and qualifications of the entir facility staff at least once per 12 months.
C. The resul ts of actions taken to co r rec t def i c i enci es occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once,per 6 months.
The per-.ormance o= activii es required by ihe Oual
'Program to meet the criteria of Aopendix "8", 10 CFRty Assurance
=0, at le s-
. once per 24 months.
e. The Facility Emergency Plan and implementing procedures at least once per 24 months.
The Facility Security Plan and implementing procedures i
'east once per 24 months.
a
g. Any other area of facility operation considered appropriate by the.HSDRC.
The Facility Fire Prot ction Program and implementing procedures at least once per 24 months.
An independent fire proteciion and loss prevention orogram inspection and audit shall be performed at least once per IZ months'utilizing ei her qualified offsite licensee personnel or an outside fire protection firm.
An inspection and audit of the fire prctection and loss pre-vention program shall be performed by a qualified outside fire consultant at least once per 36 months.
The radiological environmental moniioring program and ihe results thereof at 1eas once per 12 months.
The OFFSITE DOSE CALCULATION )ANNUAL and implemeniing procedures at leas i once per 24 months.
D. C. COOK - UNIT 2 6-11
ADMINISTRATIVE CONTROLS
m. The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at le'ast once per 24 months.
n. The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 1.21, Rev. 1, June 1974 and Regulatory Guide 4.1, Rev. 1, April 1975 at least once per 12 months.
AUTHORITY 6.5.2.9 The NSDRC shall report to and advise the Vice Chairman, Engineering and Construction, AEPSC, on those areas of responsibility specified in S'ection 6,5.2,7 and 6.5,2.8, RECORDS 6.5.2.10 Records of the NSDRC activities shall be prepared, approved and distributed as indicated below:
a. Minutes of each NSDRC meeting shall be prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days following each meeting.
b, Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days following completion of the review.
c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice Chairman, Engineering and Construction, AEPSC, and to the management positions responsible for the areas audited within 30 days after completion of the audit.
6.6'EPORTABLE 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.
I b, Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission, shall be reviewed, by PNSRC and submitted to the NSDRC Chairman.
D. C. COOK - UNIT 2'- 12
ADMINISTRATIVE CONTROLS .
6.7 SAFETY L'IMIT VIOLATION 6.7.1 The following actions shall be taken in the event a. Safety Limit is violated:
a. The facility shall be placed in at least HOT STANDBY within one hour.
b. The Safety Limit violation shall be reported to the Commission and to the Chairman of the NSDRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PNSRC. This report shall describe (1) the applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
d. 'The Safety Limit Violation Report shall be submitted to the Commission, the Chairman of the NSDRC, and the Executive Vice President-Construction and New York Engineering AEPSC within 14 days of the violation.
'6.8 'ROCEDURES 6,8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
a. The applicable procedures recommended in Appendix'"A" of Regulatory Guide 1.33, November 1972.
Refueling operations.
C ~ Surveillance and test activities of safety related equipment.
d. Security Plan implementation.
e. Emergency Plan implementation.
f. Fire Protection Program implementation.
go PROCESS CONTROL PROGRAM implementation.
h. OFFSITE DOSE CALCULATION MANUAL implementation.
Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 1.21, Rev.l, June 1974 and Regulatory Guide 4.1, Rev. 1, April 1975.
6.8,2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the PNSRC and approved by the Plant Manager prior to implementation and reviewed periodically as set forth in administrative procedures.
D. C. COOK - UNIT 2 6-13
ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 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.
6.9.1.7 The annual radiological environmental operating reports shall include summaries, interpretations, and statistical evaluation 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 surveil-lance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the land use censuses required by Specification 3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.
The annual radiological environmental operating reports shall include summarized and tabulated results in the format of 3 12-2 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 report shall be submitted noting and explaining one reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following: a summary description of the radiological environmental monitoring program including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equip-ment used; a map of all sample locations keyed to a table giving distances and directions from one reactor; the result of land use census required by the Specification 3.12.2; and the results of participation in the Interlaboratory Comparison Program. required by Specsfication 3.12.3.
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 specifiy the releases of radioactive material for each unit.
D. C. C00'K - UNIT 2 6-16
ADMINISTRATIVE CONTROLS
@I SEMI-ANNUAL 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 6Q days after January 1 and July 1 of each year.
6.9.1.9 The radioactive effluent'elease reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units as outlined in Regulatory Guide 1.21, "Measuring, Evaluating and Reporting in Solid 'Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants," with data sumnarized on a quarterly basis following the format of Appendix 8, thereof.
The radioactive effluent release report to be submitted 60 days after January 1 and July 1 of each year shall include a quarterly summary of hourly meteorogical data collected during the repor ting period. This summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction and atmospheric stability.. The report submitted 60 days after January 1 shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. These reports shall include as assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figure 5.1-3) during the reporti;ng period. All assumptions 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 determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM).
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 specify the releases of radioactive material from each unit.
D. C. COOK - UNI'T 2 6-17
ADMINSTRAT IVE CONTROLS The radioactive effluent release 'report to be submitted 6Q days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.
he radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report period:
a. Volume (cubic meters),
b. Total curie quantity (specify whether determined by measurement or estimate),
c. Principal radionuclides (specify whether determined by measurement or estimate),
=
d. Type of waste (e.g;, spent resin, compacted dry waste, evaporator bottoms),
e. Type of container (e.g., LSA, Type A, Type B, Large guantity),
and
f. Solidification agent (e.g., cement).
The radioactive effluent release report shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluent on a quarterly basis.
The radioactive effluent release reports shall include any change to the PROCESS CONTROL PROGRAM (PCP) and the OFFSITE DOSE CALCULATION MANUAL (QDCM) made during the reporting period.
MONTHLY REACTOR OPERATING REPORT 6.9.'l.i0 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office Qf Management and Program Analysis, U.S. Nuclear Regulatory Comnission, l<ashington, 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.
REPORTABLE OCCURRENCES 6.9.1.11 The REPORTABLE OCCURRENCES of Specification 6.9.1.12 and 6.9.1.13 below, including corrective actions and measures to prevent recurrence,, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution 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.
0. C. COO)( - UNIT 2 6-18
ADMIMISTRATI YE CONTROLS PROMPT NOTIFICATION WITH WRITTEN FOLLONUP 6.9.1.12 The types of events listed below shall be reported within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone and confirmed by telegraph, mailgram or facsimile transmission to the Oirector 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 vent report form shall be supplemented, as. needed, by additional narrative aterial to provide complete explanation of the circumstances surrounding the vent.
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 monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.
b. Operation of the unit or affected systems when 'any parameter or operation 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 operation greater than or equal to 1/ak/k; a calculated reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the technica1 specification; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more xnan 0.5@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 requirements of system(s) used to cope with accidents analyzed in the SAR.
f. Personnel error or procedural inadequancy. which prevents or could prevent, by itself, the fu1fillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.
0. C. COOK - UNIT. 2 6-19
ADMINISTRATIVE CONTROLS
g. Conditions arising from natural or man-made events that, as a direct result of the event*require unit shutdown, operation of safety systems, or other protective measures required by technical speci fi ca ti ons.
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 analysis in the safety analysis report or technical specification 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 pr event the existence or development of an unsafe condition;.
j. Offsite releases of 'radioactive material 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 with-in the specified limits.
D. C. COOK - UNIT 2 6-20
ADMINISTRATI VE CONTROLS THIRTY DAY WRITTEN REPORTS 6.9.1.13 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occurence of the event. The written report sha11 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 explana-tion of the circumstances surrounding the event.
a. Reactor protection system or engineered safety feature in-strument 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 re-quired by a limiting condition for operation.
c. Observed inadequacies in the implementation of administrative or procedural 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 than those specified in 6.9.1.12.c above designed to contain radioactive'aterial resulting from the fission process.
e. An unplanned offsite release of 1) more than 1 curie of radio-active material in liquid effluents, 2) more than 150 curies of noble gas. in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information:
l. A description of the event and equipment involved.
2. Causes(s) for the unplanned release.
3. Actions taken to prevent recurrence.
4. Consequences of the unpl'armed release.
f. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.12-2 when averaged over any calendar quarter sampling period. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
D. C. COOK - UNI'T 2 6-21
ADMINISTRATIVE CONTROLS PECIAL REPORTS
.9.2 Special reports shall be submitted to the Director of the ffice of Inspection and Enforcement Regional Office within the time eriod specified for each report. These reports shall be submitted overing the activities identified below pursuant to the requirements of the applicable reference specification:
a. ECCS Actuation, Specifications 3.5.2 and 3.5.3.
Inoperable Seismic Monitoring Instrumentation, Unit No. 1, Specification 3.3.3.3..
C~ Inoperable Meteorological Monitoring Instrumentation, Unit No.
1, Specification 3.3.3.4.
d. Fire Detection Instrumentation, Specification 3.3.3.8.
e. Fire Suppression Systems, Specifications, 3.7.9.1, 3.7.9,2, 3.7.9.3 and 3.7.9o4.
f. Seismic Event Analysis, Specification 4.3'3.3.2.
~ C. COOK UNIT 2 6-22
ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
a~ Records and logs of unit operation covering 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 Operating Procedures.
f. Records of sealed source and fission 'detection leak tests and results.
go Records of annual physical inventory of all sealed source material on record.
6.10.2 The following records shall be retained for the duration of the Facility Operating License:
a. Records 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.
Co Records of radiation exposure for all individuals entering radiation control areas.
d. Records of gaseous and liquid radioactive material released to the environs.
e. Records of transient or operational cycles for those facility components identified in Table 5 .7-1.
~
Records of reactor tests and experiments.
g ~ Records of training and qualification for current members of the Plant staff.
h. Records of in-service inspections performed pursuant to these Technical Specifications.
Records of Quality Assurance activities required by the QA Manual.
Records of reviews performed for changes made to procedures or equipment or review of tests and experiments pursuant to 10 CFR 50.59.
k. Records of meetings of the PNSRC and the NSDRC.
l. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13.
Records of radioactive shipments.
D. C. COOK - UNIT 2 6-23
ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent 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 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 1000 mrem/hr or less shall be barricaded 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 provided with or accompanied by one or more of the following:
a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
b. A radiation monitoring device which continuously 'integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
C~ An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.
This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit.
6.12.2 The requirements of 6.12.1, above, shall also apply to each high 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 Supervisor on duty and/or the Plant Health Physicist.
Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provid'ed they comply with approved radiation protection procedures for entry into high radiation areas.
'D. C. COOK - UNIT 2 6-24
ADMINISTRATIVE CONTROLS 6.13 ENVIRONMENTAL UALIFICATION 6.13.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of: Division of Operating Reactors "Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License No. DPR-58 dated October 24, 1980.
6.13.2 By no later than December 1, 1980, complete and auditable .
records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with .the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained as current as equipment is replaced, further tested, or otherwise further qualified.
'D. C. COOK UNIT 2 6-25
ADMINISTRATIVE CONTROLS 6 .14 PROCESS CONTROL PROGRAM PCP
6. 14. 1 The PCP shall be approved by the Commission prior to implementation.
6; 14.2 Licensee initiated changes to'he 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 PNSRC.
2. Shall become effective upon review and acceptance by the PNSRC.
6.15 OFFSITE DOSE CALCULATION MANUAL ODCM) 6.15.1. The ODOM 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 Semi-Annual Radioactive Effluent Release Report in the next report after the report period the 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 ODCH to be changed with each page numbered and provided with an approval and date box, together with appropr iat analyses or evaluations justifying the change (s);
b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
c. Documentation of the fact that the change has been reviewed and found acceptable by the PNSRC.
D. C. COOK - UNIT 2 6-26
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ADMINISTRATIVE CONTROLS g ~ An estimate of the exposure to plant operating personnel as a result of the change; and Documentation of the fact that the change was reviewed and found acceptable by the PNSRC.
2. Shall become effective upon review and acceptance by the PNSRC.
6.16.2 Commission initiated changes:
l. The applicability of the change to the facility shall be determined by the (PNSRC) after consideration of the facility design.
2. The licensee shall provide the Commission with written notification of its determination of applicability including any necessary revisions to reflect facility design.
D. C. COOK - UNI'T 2 6-28
ATTACHMENT 3 TO AEP:NRC:0055F

ML17334A428

Text

SUBJECT:

Dear Mr. Denton:

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