ML18065A441
| ML18065A441 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 01/18/1996 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18065A439 | List: |
| References | |
| NUDOCS 9601250088 | |
| Download: ML18065A441 (18) | |
Text
ATTACHMENT 1 CONSUMERS POWER COMPANY PAL ISADES PLANT -
DOCKET 50-255 TECHNICAL SPECIFICATION CHANGE REQUEST
- ADMINISTRATIVE CONTROLS ADDITIONAL CHANGES Replacement Proposed Pages 6 Pages
~-- -960f250088960118 ________ - ---,
.!SADES PLANT TECHNICAL SPECIFllIONS TABLE OF CONTENTS SECTION DESCRIPTION PAGE NO 5.0 DESIGN FEATURES 5-1 5 .1 SITE 5-1 5.2 CONTAINMENT DESIGN FEATURES 5-1 5.2.1 Containment Structures 5-1 5.2.2 Penetrations 5-2 5.2.3 Containment Structure Cooling Systems 5-2 5.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) 5-2 5.3.1 Primary Coolant System 5-2 5.3.2 Reactor Core and Control 5-3 5.3.3 Emergency Core Cooling System 5-3 5.4 FUEL STORAGE 5-4 5.4.1 New Fuel Storage 5-4 5.4.2 Spent Fuel Storage 5-4a Figure 5-1 Site Environment TLD Stations 5-5 6.0 ADMINISTRATIVE CONTROLS 6-1 6 .1 RES PONS IB IL ITV 6-1 6.2 ORGANIZATION 6-1 6.2.1 Onsite and Offsite Organizations 6-1 6.2.2 Plant Staff 6-2 6.3 PLANT STAFF QUALIFICATIONS 6-3 6.4 PROCEDURES 6-4 6.5 PROGRAMS AND MANUALS 6-5 6.5.1 Offsite Dose Calculation Manual 6-5 6.5.2 Primary Coolant Sources Outside Containment 6-6 6.5.3 Post Accident Sampling Program 6-6
. 6.5.4 Radioactive Effluent Controls Progr~m 6-7 6.5.5 Containment Structural Integrity Surv. Program 6-8 6.5.6 Primary Coolant Pump Flywheel Surv. Program 6-8 6.5.7 Inservice Inspection and Testing Program 6-8 6.5.8 Steam Generator Tube Surveillance Program 6-9 6.5.9 Secondary Water Chemistry Program 6-14 6.5.10 Ventilation Filter Testing Program 6-15 6.5.11 Reserved 6-16 6.5.12 Technical Specifi~ation Bases Control Program 6-16 6.5.13 Reserved 6-17 6.5.14 Containment Leak Rate Testing Program 6-17 6.6 REPORTING REQUIREMENTS 6-18 6.6.1 Occupati~nal Radiation Exposure Repdrt 6-18 6.6.2 Radiological Environmental Operating Report 6-18 6.6.3 Radioactive Effluent Release Report 6-18 6.6.4 Monthly Operating Report 6-18 6.6.5 Core O~erating Limits Report 6-19 6.6.6 Reserved 6-20 6.6.7 Accident Monitoring Instrument Report 6-20 6.6.8 Containment Structural Integrity Surveillance Report 6-21 6.6.9 Steam Generator Tube Surveillance Report 6-21 6.7 HIGH RADIATION AREA 6-22 iv Amendment.No.
4.4 4.5
.Deleted CONTAINMENT TESTS 4.5.1 Integrated Leakage Rate Tests The containment integrated leak rate testing shall be performed in accordance with the Containment Leak Rate Testing Program.
4.5.2 Local Leak Detection Tests
- a. Test (1) Local leak rate tests shall be performed ~t a pressure of not less than 55 psig.
(2) Local leak rate tests for checking air lock door seals within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of each door opening shall be performed at a pressure of not less than 10 psig.
(3) Acceptable methods of testing are halogen gas detection, soap bubble, pressure decay, or equivalent.
(4) The local leak rate shall be measured for each of the following components:
(a) Containment penetrations that employ resilient seal gaskets, sealant compounds, or bellows.
(b) Air lock and equipment door seals.
(c) Fuel transfer tube.
(d) Isolation valves on the testable fluid systems' lines penetrating the containment.
(e) Other containment components which require leak repair in order to meet the acceptance criterion for* any integrated leak rate test.
- b. Acceptance Criteria (1) The total leakage from all penetrations and isolation valves shall not exceed 0.60 L8 *
(2) The leakage for an air lock door seal test shall not exceed
- 0. 023 L0
- 4-19 Amendment No. -l-2-, ~' 3-S,
4.5 Basis CONTAINMEN~STS (continued)
The containment is designed for an accident pressure of 55 psig. 111 While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a temperature of about 104°F. With these initial conditions, following a LOCA, the temperature of the steam-air mixture at the peak accident pressure of 55 psig is 283°F.
Prior to initial operation, the containment was strength-tested at 63 psig and then leak rate tested. The design objective of this preoperational leak rate test was established as 0.1% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 55 psig. This leakage rate is consistent with the construction of the containment, 121 which is equipped with independent leak-testable penetrations and contains.channels over all unaccessible containment liner welds, which were independently leak-tested during construction.
Accident analyses have been performed on the basis of a leakage rate of 0.1% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With this leakage rate and with a reactor power level of 2530 MWt, the potential public exposure would be below 10 CFR 100 guideline values in the event of the Maximum Hypothetical Accident. 131 .
The performance of a periodic integrated leak rate test during plant life provides a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment.
In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic leak rate test is to be performed without preliminary leak detection surveys or leak repairs and containment isolation valves are to be closed in the normal manner.
This normal manner is a coincident two~of-four high radiation or two-of-four high containment pressure signals which wil.l close all containment isolation valves not required for engineered safety features except the component cooling lines' valves which are closed by CHP only. The control system is designed on a two-channel (right and left) concept with redundancy and physical seoaration. Each chann~l is capable of initiating containment isolation. 141 The Type A test requirements including pretest test methods, test pressure, acceptance criteria, and reporting requirements are in accordance with the Containment Leak Rate Testing Program.
The frequency of the periodic integrated leak rate test is keyed to the refueling schedule for the reactor because these tests can best be performed during refueling shutdowns. The specified frequency is based on three major
- considerations. First is the low probability of leaks in the liner because
- of (a) the test of the leak tightness of the welds during erection; (b) conformance . of the complete containment to a low leak rate at 55 psi g during preoperational testing which in consistent with 0.1% leakage at design basis accident (OBA) conditions: and (c) absence of any significant stresses in the liner during reactor operation.
4-22
.Amendment No. -l-99, ~
4.5 CONTAINMENT TESTS Basis (continued)
Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.60L of the total leakage that is specified as acceptable from 8
)
penetrations and isolation valves. Third is the Containment Structural Integrity Surveillance Program which provides assurance that an important part, of the structural integrity of the containment is maintained.
The basis for specification of a total leakage rate of 0.60 La from penetrations and isolation valves is specified to provide assurance that the integrated leak rate would remain within the specified limits during the intervals between integrated leak rate tests. This value allows for possible deterioration in the intervals between tests.
The basis for specification of an airlock door seal leakage rate of 0.023 La is to provide assurance that the failure of a single airlock door will not result in the total containment leakage exceeding 0.6 La. The seven (7) day LCO specified for exceeding the airlock door leakage limit is acceptable since it requires that the total containment leakage limit is not exceeded.
References (1) Updated FSAR Section 5.8.2.
(2) Updated FSAR Section 5.8.8 (3) Updated FSAR 14.22 (4) Updated FSAR Section 8.5.1.2 (5) 10 CFR Part 50, Appendix J.
(6) Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program", September 1995.
4-23 Amendment No. 2-, -l-G9-, 2-6, 3-5,
6.0 6.5.13 ADMINISTRA~ CONTROLS Reserved 6.5.14 Containment Leak Rate Testing Program Programs shall be established to implement the leak rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option 8, as modified by approved exemptions. The Type A test program shall meet the requirements ~f 10 CFR 50, Appendix J, Option B and shall be in accordance with the guidelines of Regulatory Guide 1.163, "Performance-Based Containment Leakage-Test Program, dated September 1995." The Type Band Type C test program shall meet the requirements of 10 CFR 50, Appendix J, Option A, as modified by the exemption from certain requirements of 10 CFR 50 Appendix J which were granted in an NRC letter to Consumers Power Company dated December 6, 1989.
The containment is designed for an accident pressure, Pa, of 55 psig; the maximum allowable containment leak rate, La, at Pa, shall be 0.1% of containment air weight per day.
Leak rate acceptance criteria are:
- a. Containment leak rate acceptance criteria is s 1.0 La. During the first plant startup following testing in accordance with this program, the leak rate acceptance criteria are s 0.60 La for the Type B and Type C tests and s 0.75 La for Type A tests;
- b. Air lock testing acceptance criteria are:
- 1) Overall air lock leak rate is s 0.60 La when tested at ~ Pa;
- 2) For each door, leak rate is s 0.023 La when pressurized to
~ 10 psig.
The Surveillance interval extensions of LCO 4.0.2.a and 4.0.2.b are not applicable to the Containment Leak Rate Testing Program requirements.
The provisions of LCO 4.0.3 are applicable to the Containment Leak Rate Testing Program requirements. *
- 6-17 Amendment No.
6.0 6.7 AIJMINISTRA~ CONTROLS HIGH RADIATION AREA 6.7.1 Pursuant to *10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiatton area, as defined in 10 CFR 20, in which the intensity of radiation is > 100 mrem/hr but < 1000 mrem/hr, 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 (RWP). Individuals qualified in radiation protection procedures (e.g., health physics technicians) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates < 1000 mrem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.
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 that continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device that 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 levels in the area have been established and personnel are aware of them.
- c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for.
providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Work Request.
6.7.2 In addition to the requirements of Specification 6.7.1, except as allowed by 6.7.3, areas with radiation levels ~ 1000 mrem/hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.
6.7.3 For' individual high radiation areas with radiation levels of ~ 1000 mrem/hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.
6-22 Amendment No. 48, MS, -l-64,
ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 TECHNICAL SPECIFICATION CHANGE REQUEST ADMINISTRATIVE CONTROLS ADDITIONAL CHANGES Existing Pages Marked to Show Changes 10 Pages
4.* 5 CONTAINMEN~STS
- At>Bl i cabi l i ty Applies te ceRtaiRmeRt leakage aRd structural iRtegrity.
Objective Te *1eri fy that peteRti al leakage frem the ceRtai RmeRt aRd the prestressi Rg teRdeR leads are mai Rtai Red ~ii thi R specified values.
Sf>ecificatieRs 4.5.1 Integrated Leakage Rate Tests A surveillaRce test pregram fer the containment everall integrated
~
ChaRge Ne. 16 Amendment No. 2-, 3-§.
4.5 The Type A test requirements including pretest test methods, test pressure, acceptance criteria, and reporting requirements are in
- ~~~~~:~~: iiit:P:ii;,i:l;imili:::::::iliii:lii1]m1ii:liiiiiiilim~r appre\*ed The frequency of the periodic integrated leak rate test is keyed to the refueling schedule for the reactor because these tests can best be performed during refueling shutdowns. The specified frequency is -a-s-specified iFI IQ CFR Part §Q, AppeF1dix J which is based on three major considerations. First is the low probability of leaks in the liner because of (a) the test of the leak tightness of the welds during erection; (b) conformance of the complete containment to a low leak rate at 55 psig during preoperational testing which in consistent with 0.13 leakage at design basis accident (DBA) conditions: and (c) absence of any significant stresses in the liner during reactor operation. Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.60La) of the total leakage that is specified as accepta~J~ ..,.ft.C>.R.l ... P.~~~.~r~.t..i.~~.~ ....~.!'!.~: .. ..i.~~J.~.t.i on valves. Thi rd is the teF1deF1
~;~~~~e,l!l~,l~~,~~l!:~[~~i!ilf;i,JiililHi!!l!Mlli.i survei 11 ance program which Amendment No. 12-, 09, -l-3-&
4.* 5 CONTAINMEN-~STS (Cont'd]
CeRtaiRmeRt dame delamiRatieR iRspeetieRs perfermea iR 1979 aRa 1982 have eeRfirmed that Re eeRerete aelamiRatieR ha~ eeeijrrea. The pessihility that aelamiRatieR might eeeijr iR the fijtijre is remete heeaijse dame teRaeR prestress ferees graaijally dimiRish threijgh ~ormal teRdeR relaxatieR aRa eeRerete streRgth Rermally iRereases ever time.
Te aeeeijRt fer this remete pessihility, he~1ever, aR adaitieRal del ami Rati eR i Rspeeti eR wi 11 he performed .i R the eveRt that 51' or more ef the iRstalled teRdeRs mijst he reteRsieRea ta eempeRsate for excessive less ef prestress. This i~speetieR weijld he ta eeRfirm that aRy systematie exeessive prestress loss did Rat resijlt from aelamiRatieR aRa that the reteRsioRiRg precess die Rat resijlt iR aelamiRatieR.
References (1) FSAR, SectieR 5.1.21 Updated FSAR section 5.8.2.
(2) FSAR, SectioR 5.1.8; Updated FSAR section 5.8.8 (3) FSAR aRd Updated FSAR 14.22 (4) FSARr SectioR 8.5.4; Updated FSAR Section 8.5.1.2
{5) FSAR aRd IJpaated FSAR SectieR 6.2.3
{6) FSAR, SectieR 5.1.8.4; FSAR, AmeRameRt No. 14, QijestieR §.37; aAd IJpdated FSAR SectieR 5.&.8.3.
{7) IJpdated FSAR, SectieR 5,8.8.6
<8=1> 10 CFR Part 50, Appendix J.
Amendment No. ~, -1{}9
ADMINISTRATIVE t:
CONTROLS~
- 0 1
~:~=~~~ i~~ ~:;~1c:at~~e~a:~ !~i h~~r:~f~!t~Ad mechaAical SAijbbers cevered by SpecificatieA 3.20. This shall iAClijde the date at which the service life cemmeAces aAd asseciated iAstallatieR aAd maiAteAaRce recerds.
- m. Recerds ef traiAiAg aRd qijalificatieAs fer members ef the plaAt staff.
A. Recerds ef reacter tests aRd experimeRts.
- e. Recerds ef reviews perfermed fer chaAges made te the OFFSITE DOSE CALCULATION MANUAL aAd the PROCESS CONTROL PROGRAM.
6.11 RADIATION PROTECTION PROGRAM
- =~~:===i !t~hpi~: ~~~!1~:~!=i! :fprgtg~~i~~,s==~ st:1~rg~a=:~reved, 0 0 1 maiRtaiAed aAd adhered te fer all eperatieAs iAvelviAg perseAAel radiati~R expesijre.
6.12 HIGH RADIATION AREA 6.12.1 IA lieij ef .the "ceAtrel device" Bf "alarm sigAal" feqijired by 10 CFR 20.203(c)(2), each high fadiatieA area iA which the iAteAsity ef radiatieA is greater thaA 100 mRem/hf bijt less thaA 1000 mRem/hr shall
~=t~:~=!c;~:~e;:ds~:~~P~~ij:==i~ef~:jegya;e:ij~!T~grt:!:!!:~ :fe; aAd RadiatieR Werk Permit.*
HIGH RADIATION AREA 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.
Amendment No. le, 48, ~, -l-98, .J.S4
ADMINISTRATIVE CONTt::
c.
An individual qualified in radiation protection procedures who is eql:Ji1313ee:I with a radiation dose* rate m.9.ni...t.P..r..t..n.9. . . .Q.gy.J..~-~--~. . . .*.*.Ih.i...§.
~~~!~~1~~ 1 p~~nlv~e c~~~~~~s~~!~ i~= :!12,~e::~:1'T=~1~11w~'~T=:~:~,R:!:11µea and shall perform periodic radiation surveillance at the frequency specified by the Radiation Work Permit.
6.12.2
!~tehb:h~ r=~=~=t~A:f ;~~1!tt!=ei;J3;!~ai:re:h~Ah~~8gr:~::;~:=r~r¥= iA ae:le:liti eA, l eekee:I Eloors shall be 13rev1 e:leEI te 13reveAt l:JAal:Jtheri zeEI eAtry iAte sl:Jeh areas (> 1999 111remfhel:Jr) aAe:I the keys shall be 111aiAtaiAeEI l:JAe:ler the ae:lmiAistrative eeAtrel ef the Shift Sl:J13ervisor &A ell:Jty aAEl/er the PlaAt Health Physicist raEliatieA safety sl:J13erviser.
Amendment No. 4&
4_.5
. CONTAINME~~STS Aeel i cabi l i t*1 Applies te ceRtai RmeRt leakage aREI. stn1ct1:rl"al i Rtegri ty.
Objecti'+'e Te '+'erify that peteRtial leakage frem the ceRtaiRmeRt aREI the prestressiRg teAEleR leaels are maiRtaiReel ~~ithiR specifieel 't'alijes.
SeecificatieRs 4.5.1 Integrated LeakaEJe Rate Tests A Sijr'+'eillaRce test pregram fer the containment e't'erall integrated
~
ChaRge Ne. 16 Amendment No . .f-2., 3-S
4.5 The Type A test requirements including pretest test methods, test pressure, acceptance criteria, and reporting requirements are in
- ~~~~~:~~: ili~:~[l1i:ii1ii11::illi[:l.ii1~:mvii:li\iiiiiiim~r appre*ied The frequency* of the periodic integrated leak rate test is keyed to the refueling schedule for the reactor because these tests can best be performed during refueling shutdowns. The specified frequency is vs-speeifiee iR 19 CFR Part 59, AppeRdix J ~1hieh is based on three major considerations. First is the low probability of leaks in the liner because of (a) the test of the leak tightness of the welds during erection; (b) conformance of the complete containment to a low leak rate at 55 psig during preoperational testing which in consistent with 0.1%
leakage at design basis accident (OBA) conditions: and (c) absence of any significant stresses in the liner during reactor operation. Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.60L of the total leakage that is specified as 8 )
- ~~:~;a~~bl,ithmihiit:si£ai~fi=Fimi:t:~t},,,~:F:l=ii~:i ~~r~=~ ~~~~ce Th~~d r!~ !~~ c~eRdeR Prov i de's;,,,,,,a's' stJ'Fah'C'e' ' ' l'ha't' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' 'n,,,,::J,,,,,,,M P g Amendment No. 2-, .J..G9, +/-3-9
CONTAINMEN.ESTS {Cont'd)
CeRtaiRmeRt dame delamiRatieR iRspectieRs perfermea iR 1970 aAd 1982 have ceRfirmed that Ae ceAcrete delamiRatieA has eccijrred. The pessibility that delamiRatien might eccijr iR the fijtijre 1s remete becaijse dame teRdeA prestress farces gradijally dimiRish threijgh Rermal teRdeR relaxatieR aRd ceRcrete streAgth Aermally iAcreases ever time.
Te acceijRt fer this remete pessibility, heue*1er, aR additieRal delamiRatieR inspectieR will be perfermed iR the eveRt that 5% er more ef the iRstalled teAdeAs mijst be reteRsieRed te cempeRsate fer excessive less ef prestress. This inspectien weijld be to cenfirm that aRy systematic excessive prestress less did Aet resijlt frem delaminatieA aAd that the reteAsieAiAg precess did Aet resijlt in delamiAatieA.
References
{l) FSAR, SectieA 5.1.2; Updated FSAR section 5.8.2.
(2) FSAR, SectieA 5.1.8; Updated FSAR section 5.8.8 (3) FSAR aAd Updated FSAR 14.22 (4) FSAR, SectieA 8.5.4; Updated FSAR Section 8.5.1.2 (5) FSAR aAd Updated FSAR SectieA 6.2.3 (6) FSAR, SectieA 5.1.8.4; FSAR, AmeAdmeAt Ne. 14, QijestieA 5.37; aRd Updated FSAR SectieA 5.8.8.3.
(7) Updated FSAR, SectieA 5,8.8.6 CS,!> 10 CFR Part 50, Appendix J.
Amendment No. ~' -l-G ADMINISTRATIVE t:
CONTROLS*~ **
~~:=~=~ :: ~h:e~~;~{e:aii~e~a=~l !~! h~=ri:~f J!t~Rd meehaR i eal SRHbberis eeveried by SpeeifieatieR 3.20. This shall iRelHde the date at ~~hi eh the serivi ee life eemmeRees aRd asseei ated **
iRstallatieR aRd maiRteRaRee rieeerids.
- m. Reeerids ef triaiRiRg aRd ~HalifieatieRs fer> members ef the plaRt staff.
R. Reeerids ef rieaeteri tests aRd experiimeRts.
- e. Reeerids ef rieviews periferimed fer> ehaRges made te the OFFSITE DOSE CALCl:JLATION HANl:JAL aRd the PROCESS CONTROL PROGRAM.
Ei .11 RADIATION PROTECTION PROGRAM 1
- ~:;i~===~ !t;hpi~:e~~:!1~:~!:i!e:fpl~t~~~i~B,s~=~ st:1fri~~a=:~rieved, maiRtaiRed aRd adhered te fer> all eperiatieRs iRvelviRg periseRRel riadiatieR ex~esHr>e.
Ei .12 HIGH RADIATION AREA S.12.1 IR lieH ef the eeRtriel de*1iee eri alarim sigRal 11 11 11 11 rie~Hir>ed by
!~d~~:1:~-~~a~~!ii!~ ih:~ nHhmR:!)h; l!~tat~:s ~h:~i~~o~h:R!:;h~s!~~1if 1 1
~=t~:~=!e:~:~e*::d s~:n~~:H:==t~efl:~e~Y a~e:H~!i~g rii~!~!~:~ :fe: aRd RadiatieR Werik Perimit.*
HIGH RADIATION AREA Any individual or group of individuals per!llitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indi~ates 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.
Amendment No. le, 48, &9, -MS, -l.s4
ADMINISTRATIVE c.
CONTR~
- An individual qualified in radiation protection procedures whe is e~ l:i i 1313 ed wi th a rad i at i on dose rate m.QJJ i. .t.9.r.J.n9 . ...Jl..!..Y...t~.~- .~. . . . . . .I~J.$..
i Ad h i dl:i al sh a11 be 'f'e s eAs i b1e f e.,. ::=:::=t:***hnttt):wr*'::***::**:**=*:**=*:**=:**\\l)]::*****=:=tft**:**:::
prov id i ng po s i ti ve cont ~o l over the 'i'E,'i'V'f~l'~=:g=!:~~,,~l'~'*=*tR'~*=!~e a and shall perform periodic radiation surveillance at the frequency specified by the Radiation Work Permit.
6.12.2 The abeve 'f'e~l:ii'f'emeAts shall alse a1313ly te each high 'f'adiatieA area iA which the iAteAsity ef radiatieA is ~reater thaA 1000 mremjhol:ir. IA additi oA, 1 ecked deers shall be 13rev1 ded te 13reveAt l:iAal:itheri zed eAtry iAte Sl:ich areas (> 1000 mremfhel:ir) aAd the keys shall be maiAtaiAed l:iAder the admiAistrative ceAtrel ef the Shift S1:i13erviser eA dl:ity aAd/er the PlaAt Health Physicist radiatieA safety Sl:i13erviser.
Amendment No. 48-