ML20127J546
| ML20127J546 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/15/1973 |
| From: | Anderson F, Ziemann D US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20127J540 | List: |
| References | |
| NUDOCS 9211190367 | |
| Download: ML20127J546 (34) | |
Text
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UNITED STATES AIONIC ENEEGY 00 W18810N SA751T TY.AJSPj* d BY THE DIRsC51* ATE OF LICENSING liORTIEPJ JTAW. N,WER COMPANY _
pocket 100.'50-263 CRANGE NO. 12 TO TECHNICAL SPECIFICATIONS By latters dated October 26 sad October 31, 1973. Northern States Power Company (NSP) proposed changes to the Technical Specifications of Provisional Operating License No. DPR-22 for the Monticello Nuclear Generating Plant that would correct errors and inadequacies, correct discrepancies in Change No. 2 to the Technical Specifications dated January 14, 1972, and incorporate revisions to the Technical Specifi-cations necessary for the operation-of the augmented off-gas mystem.
Wu have reviewed the proposed changes and the reasons for the changes-submitted by NSP. Our evaluation and discussion of these changes is presented in th6 order of the Technical Specifications pages to be changed. ,
s- ~
We compared the proposed change co. :he fuel cladding integrity safety lhait (Specification 2.1.B) of 5 percent design core flow with other BWR plants of similar design, .uch as Vertout Yankee. Ne concluded that the proposed change is acceptable and is consistent with.the requirements of other BWR plants.
We reviewed the proposed addition to Tabis 3.1.1 of a Limiting Trip Setting on the APRM downscale trip. The proposed setting should be " greater than or equal to 3/125 of full scale" rather than "less than or equal to 3/125-of full scale" as proposed. With-this modifi-cation, the- trip setting is consistent with the function of the APRM and is an acceptable trip setting limit. The NSP staff has been consulted and has agreed with this lascessagandification.
~
The changestto be incorporated into the Technical Specifications-by Change No. 2 dated January 14, 1972,;have been compared with the
- proposed changes incorporated herein. We have-concluded that Chance No. 2, which was approved to be effective following installation of omct > . . . . . . . . . . . . . . . . . . ..- ..... . . . . . . . . . . . . . . . . . . . . . . . . . .
$URNAME> . _ . . ... . . . . . . . , .... ... . . . . . . . . . .
DATE >
- **3-to-s t**5-1 +4s-ers Form AEC-lis (Rev.9-33) AEChi 0240 9211190367 731115 PDR- ADOCK 05000263 P PDR-
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l , t the augmented off-gas system, should be deleted in its entirety.
The changes proposed by the October 31, 1973 substittal and modified by us should be incorporated into the Technical Specifications to replace those changes approved for incorporationby Change No. 2.
These changes will be consistent with current Regulatory require-ments for the augmented off-gas system.
We have reviewed the proposed changes to Specifications 3.2.D. 3.2.E.2, and Table 4.2.1, including Bases, regarding tha radiation monitoring requirements and set points for the steam jet air ejector monitors and the reactor building veut monitors. We have modified the limit set point (Specification 3.2.D.1) for the air ejector monitors to reflect the AEC staff's changes to Specification 3.8.A.1 on gaseous effluent limits. The proposed lowering of the trip set point for the ventilation plenum is consistent with the gaseous affluent limit for the reactor building vent releases as determined by the \EC staff to meet 10 CFR Part 2011Juits. The proposed change to Table 4.2.1 clarifies the calibration method to be used. The Bases have been modified to reflect our changes and justify the limits set by the proposed changes.
Tables 3.2.1 and 3.2.5 regarding HPCI High Steam Flow trip setting and time delay setting and deviations to the APRM and RBM trip settings have been revised to clarify the operation of the existing trip system.
We have reviewed the proposed changes to Figures 3.4.1 and 3.4.2 regarding liquid poison volume concentration requirements. We have concluded that the proposed changes are consistent with current Ragulatory requirements and identify the chemical form of the liquid poison.
The changes to Specifications 4.6.D and 4.7.C.1.a-c with Bases which delete reporting requirements that have been fulfilled and add reporting requirements on system surveillance are appropriate and necessary to meet Regulatory requirements. Other minor changes to
_ these technical specifications are necessary for clarification.
We have reviewed the proposed changes to Specifications 3.8 and 4.8 regarding radioactive effluents,- except for the liquid effluents.
Although we agreed with the intent of the proposed changes. i.e.,
to reflect as low as practical limits ror the airborne effluents with the augmented off-gas system operating, we made major modifi-cations to the technical specifications to include current Regulatory I
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SURNAME > ' '
l DATE > 1 . .........f . . .
Form AEC-3ta (Rev. 9.m AECM 0240 ao . u- m --m 4 ass us%n
, _ _ - . - . - - ~ ~ ~ . ~ ~ - ~~
,, , _, _ , , ,- + ---- e- -+e-
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requirements on release rates, current dose analysis teethods, AEC staff tastoorological models, and current surveillance requirements.
The Bases for the technical specifications relative to these inodifi-cations have been revised completely by the AEC staff. NSP repre-sentativos have agreed with these snodifications.
The proposed change to Specification 6.7.A.2.1 is consistent with current Regulatory requirements for reporting of occupational personnel
- radiation exposure and, therefore, is acceptable.
On the basis of our evaluation, we have concluded that the proposed changes, as modified, do not present significant hasards considerations and that there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner.
, The Technical Specifications should be changed as proposed by NSP and modified by the AEC staff.
J 1
Fredric D. Anderson
. Operating Raactors Branch #2 Directorate of Licensing Ya Dennis L. Ziemann, Chief Operating Reactors Branch r2 Directorate of Licensing
...,, . _ .4 ,
. Date: 'n- 3/ a F
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e IIOTES
- 1. RATED TUJER 13 1t7/0 MWt .
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- 2. DESIGl F10W IS $7.6 x I M' y g37;33
- 3. TC/ PAL PEAKIllG FACPOR IS$ 3 08 16 CORE PRESSURE 13 1 600 P0lG 1 N 2"i10
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- 5. WATEh LEVEL In 10 FT. o III . -
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TABLE 3 1.1 .
REACIDR PROTECTION SYSTEM (SCHAM) INSTHIIMENT REQlIIREMENTS Modes in which func- Total No. of Min. No. of Operable Limiting tion must be Oper- Instrument or Operating Instru-Trip-Settings able or OperatinF** Charnels per ment Channels Per Required frip Function Refuel (3) Startup Hun Trip System Trip System (1) Condition
- 1. Mode Switch in Shutdown x x x 1 1 A
- 2. Manual Scram x x x 1 1 A 3 Neutron Flux IRM i120/125 (See Note 2) of full scale x x x(c) 4 3 A
- a. Ingh-High
- b. Inoperative 14 Flow Referenced See Specifi-Neutron Flux APRM cations (See Note 5) 2 3A.1 x 3 2 A or B
- c. High-Iligh
- b. Inoperative
- c. Downscale > 3/125 o f full scale 5 High Reactor Pressure 6 1075 psig x x(f) x(f) 2 2 A
- 6. High Drywell Pressure 6 2 poig x(1) 4 x(e, f) x(e,f) 2 2 A
- 7. Reactor Iow Water Level 3 7 in.(6) x x(f) x(f) 2 2 A
- 8. Scram Discharge Volume High Level 4-32 gal.(8) x(a) x( f) x(f) 2 2 A 9 Turbine condense -
Low Vacuum 3 23 in. I!g x(b) x(b, f) x(f) 2 2 A or c 3 1/4.1 ,o
$EV (11/15/73)
O 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS B. Emergency Core Cooling Subsystems Actuation When irradiated fuel is in the reactor vessel and the reactor water temperature is above 212 F, the limiting conditions for operation for the instrumentation which initiates the emergency core cooling sybsystems are given in Table 3.2.2.
C. Control Rod Block Actuation The limiting conditions of operation for the instrumentation that initiates control rod block are given in Table 3.2.3.
D. Air Ejector Off-Gas System
- 1. Except as specified in 3.2.U.2 and 3.2.D.3, both steam jet air ejector off-gas radiation monitors shall be operable during reactor power operation.
The trip settings for the air ejector monitors, except as specified in 3.2.D.4, shall be set to close within 30 minutes the recombiner train inlet valve (s) at a level not to exceed the equivalent of the lin.its stated in' Specification 3.8. A.1 for the off-gas stack af ter a decay time of 30 minutes.
32/h.2 LS RE'i (11/ l '; / 73) -
. -. - - - . ~ . . - . .. . . .- .. . . .
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
- 2. From and after the date that one of the two steam jet air ejector of f-gas radiation moni tors is . made or found to be inoperable, continued reactor power operation is permissible provided the inoperable radiation monitor ins t rument channel is ti!pped.
- 3. Upon loss of both steam jet air ejector off-gas radiation monitors, an orderly shut-dawn shall be initiated and the reactor shall be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 If operation is necessary with the Off-Gas !!oldup System recombiners bypassed, the steam jet air ejector radiation nouitors shall'be set to close the off-gas iso tation valve instead of the recombiner inset valves with a delay time not to exceed 15 minutes.
3.2./4.2 48A REV (11/15/73)
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Table 3.2.1 - Continued Min. No. of Operable Total No. o f Ins t ru- or Ope rat ing Instru -
rnent Channel s Per ment Channels Per Trip Required Trio Settings Trip Systen System (1.2) Conditions Function
- b. liigh Drywell Pressure 6 2 psig 2 2 0 (5) 3.. Reactor Cleanup Systen (Group 3)
- a. Low i<eactor Water _3_10'6" above the top of the 2 2 E Level active fuel
- 4. IIPCI Steam Lines 2(4) 2 F l
- a. IIPCI liigh Steam Flow 6150,000 lb/hr with $ 60 second time delay IIPCI liigh Steam Flow S .9 C,000 lb/hr 2(4) 2 F b.
16(4) 16 F
- c. IIPCI Steam Line .- 552000f I
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- 5. RCIC Steam Lines
- a. RCIC liigh Steam 645,000 lb/hr 2(4) 2 G 16(4) 16 G
- b. RCIC Ste a Line Area 6 2 GOOF liigh Temp.
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k Table 3.2.5 - Continued Trip Function and Deviations Trip Function Deviation Low-Lou Reactor ' dater Level -3 Inches Instrumentation W at Initiates Emergency Core Cooling Systems Reactor Low Pressure (Firip -10 psi Table 3.2.2 Start) Pe rmissive liigh Dryueli Pressure . +1 psi Iow ikactor Preasure (Valve -10 psi Permissive-IRIt Downscale -2/125 of Scale instrumentation Ihat Initiates. IF21 Upscale +2/125 of Scale Rod Block Table-3.2.3 -2/125 of Scale APict Downscale APRI! Upscale See Basis 2.3 - rage 24 RBM Downscale -2/125 of Scale RB!1 Upscale Same as APFli Upscale
/
A violation ui this speci fication 'is asmiu ed to occur only t. hen a device is knowingly set outside of the' limiting trip settings, or, when a sufficient number of devices have been affected by any means such that l the automatic function is incapable of operating within the allowable deviation while in a reactor mode in which the specified function must be operable or when actions specified are not initiated as specified. e t 70 l 3.2 BASES REV (11/15/73) I I
8 E S
~
I I I j l I I e 0 - S 24 -
%es a -
o 22 - 21 4% 21.4*. 2895 g31 3 1400 gal 3 20 - r z w o 18 - x w (L -
. 16 .P s
14.l* . 2210 gal ,
~
12 10.67. 2893 gal 10 i l l 1 l l 1000 2000 3000 NEI TANK VOLUME (gallons FIGilitE 3.L.1. S. liurn Pental=> rate Solutic.n Wlurne
- Concent ration Itequirennents .-
2- (l1!!5/71 3, n /;,, t,
<w
~.
m N th eg O w Q
<?.+1 {
1 I J e 150
$0LUTION TEMPERATURE MU$T BE '
EQUAL 70 0R GREATER THAN TH AT ' 140
- INDICATED BY THE CURVE i
1X) -
- i
_ 120 - w a - 3 110 - E lt' 109 - 5.-
=
52 5 90 - 8 1
. 80 3
4 i ! 70 60 - s ! $0 t ' ' ' ' 40 i S 10 'O 30 40 50 4 i nEIGHT PERCENT $0010M PENT ABORATE IN $0LUT10N (as W/o Na2 B 10 0 16
- 10110) 2 t
f a. \. 4, .w l- PIGURE 3.k2 SODIUM PENT A40R Af t SOLUTION TEMPERATURE RESU4REMENTS I l
. . - _ _ . . . _ . - - . . _ . . - _ . . _ - . . . . ~ . . . _ , . . . . . _ . . . _ . _ _ _ . . . . _ . _ _ . . _ _ . . _ _ _ . _ . _ _ . - . _ . . _ . . _ . _ _ , - . . . _ . . - . . - _ . .
t f Bases Continued 3.6 and 4.6: t I- t
- D. Coolant Leakage The former 15 gpm limit for leaks from unidentified sources was established assuraing such leakage was coming j from the primary system. Tests have been conducted which demonstrate that a relationship exists between the size
, of a crack and the probability that the crack will propagate. From the crack size a leakage rate can be detemined. For a crack size chich gives a leakage of 5 gpm, the probability of rapid propagation is less than 19-5 Thus, an l unidentified leak of' 5 gpm when assumed to be from the primary syr, tem had less than one chance in 103,000 of prepa- i gating, which provides adequate margin. A leakage of 5 gpm is detectabic and measurable. ne 21 har period I allowed for determination of leakage is also based on the lou probability of the crack propagating. The capacity of the dryuell sump pumps is 100 gpm and the capacity oi the drywell equipcent drain tank pumps- : also 100 gpa. Renioval of 25 gpm f rom either of these sumps can be accemplisited with censiderable margin. I is , An annual report will be prepared and submitted to the AEC surna rizing the primary polant tm f ryell Icakage measurements. Other techniques for detecting Icaks and the applicability of these techniques te the ".ccticello Plant will be the subject of continued study. ! i ' E. Safety and Relief Valves l s Experience. in safety valve operation shows that .1 testing of SGI of ti e safety valves pcr ref alins vutag,e is adequate to detect failures or deterioration. A tolerance value is spet ified in Section III af ic AS"E Sciler and j
- Pressure Yessel Code as +1% of the set pressure. An analysis has been performed which shcus that with all safety valves set 1% higher than the set pressure, the reactor coolant pressure safety limit of 1375 psic is not exceeded.
- j. Safety / relief valves are used to minimize activation of the safety valves. The operator will set the pressure
- settings. at ' or below the settings listed. However, the actual set points can vary as listed in the basis cf l; Specification 2.4.
- The required safety valve steam flow capacity is detemined by analyzing the pressure rise ac m panying the 1
main steam flow stoppage resulting from a MSIV closure with the reactor at 1670 twt. We ar.alysis assumes no ( 4 MSIV closure scram, but a reactor scram from indirect means (high flux). ihe relie f and safety val.ve capaci ty is assumed to total 83.97. (477 relief and 36.91 sa fety) of the full power steam generat iin rate. Bis capacity corresponds to assuming that four safety /reitef valves (477.) and four safety valves (36.97.) ope rated. i i ' 3.6/4.6 BASES j 13 '. F E'J (11/15/73)
ie \ r i !- ii.itI'li1fttl ;t l-lL ;.Ffl}{it' (j!ii li !!iiI(fI [tii- ' il* - ? e r )h c. l i ot eyw- n n a ) 3 - . . s t a ooele gl e 7
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.t 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SUiWEILLANCE REQUIRE!!ENTS --.
- 1. The maximum release rates of gross radio- l 1. Radioactive gases released from the ofi-activity shall not exceed a rate Q in gas stack and reactor building vent shall curies /sec: be continuously monitored. Station records of ot f-pis stack release rates of gross
_ E- \ y gamin radioactivity shall be maintained 93(d4_\+qD(3i4 _4 , 0.18) 0.028 0.019 ) on an hourly basis to assure that the
. spectifed rates are not being exceeded, 2.
I.ne release rates of gross radioactivity and to yield information concerning shall not exceed 16 percent of the limit .
.. general integrity of t.ne fuel cladding.
In Specification 3.8. A.1 averaged over Records of isotopic analysis shall be any calendar quarter. ma in t a ined . The off-gas stack and , reactor building vent monitoring systen
- 3. The maximum release rate of radiolodine 3 7 g ,
4 131 (I-131) shall not exceed a rate y, tu - - 1 and calitrated quarterly with an appro-microcuries/see: priate standard radiation source. Each monitor, as described, shall have a q,i e tk5 1 sensor check at least daily. 23 g32
- 4. The nele.u.c rate of I-131 sha l l not exceed 2. A steam jet air ejector of f-gas sample 4 percent of the limit in Specification shall be taken and an isotopic analysis 3.8. A.3 averaged over any calendar quarter. for at least six fission product gases; Xe-138, Xe-135, Xe-133, Kr-88, Kr-85m, ;
- 5. The maximum release rates of radioactive Kr-87 shall be made at least weekly and particulates with half-lives greater than 8 follauing each refueling or other days shall not exceed a rate Q, in micro- occurrence which could alter significanti curies /sec: the mixture of radionuclides. i l' Qi QRS 2x108 MPCa< - y i 9.5x109 HPI'a +
1 where IIELCa is t he composite ruximum per- .i I missible concentration in air in uCi/ml ) determined using Appendix B, Table II, Colunut 1 and Notes of 10 CFR 20. g [, I 3.8/4.8 169 [ REV C 1/15/ 7 3 ) f i . t
. ~ - _- . . - . - - - - . . - . . - . - - - - . - - -. .- - . - - - - . . . _ - - - - . . . ..
. .=- . _
1 j - i
^ * -
3.0 LIMITING COND1110NS FOR OPERATION 4.0 SURVEILI.ANCE REOUIRFifENTS __ __ [
- 6. The release rates of radioactive 3. Gaseous release of tritium shall be particult.tes with half-lives greater calculated on a quarterly basis from than 8 days shall not exceed 8 percent tritium concentration of the condensate.
of the limit in Specification 3.8.A.5 Vaporous tritium shall be calculated averaged over any calendar quarter. from a representative sample. The , num of these two values shall be re- i
- 7. If the caximum release rate limits of Spec- ported as the total tritium release.
I ifications 3.8.A.1, 3.8.A.3, or 3.8.A.5 are ; not met following a routine surveillance 4. Radioiodine and radioactive particulates 4 check, an' orderly shutdown shall be with hall--lives greater than 8 days initiated and the reactor shall be in the released f rom the of f-gas stack and
; cold shutdown condition within 24 hours. reactor building vent shall be continuously ,
i sampled. Station records of release of
- 8. If the limits of Specification 3.8.A.2, all radioiodine 131 and particulates
; 3.8.A.4, or 3.8.A.6 are exceeded, with half-lives greater than 8 days I
appropriate' corrective action such as an shall be maintained on the basis of all I orderly reduction of power shall be stack and vent cartridges counted. l initiated to bring the releases within The charcoal cartridges shall be counted th se limits.
. weekly when the measured release rate l' of radioiodine 131 activity is less
- 9. if the release rates exceed four percent than the rate of Specification
! of the limits in Specification 3.8.A.' 3.8.A.4; otherwise the cartridges 4
averaged over any calendar quarter e 40 shall be counted daily. The
, percent of the-limits in Specificat a particulate filters shall be counted
) 2.8. A. 3 or 3.8. A.5 averaged over an weekly when the measured release
; calendar quarter, the following actwas rate of particulate radioactivity shall be taken: with half-lives greater than S j days is less than the rate of .
I Specification 3.8.A.6; otherwise .l the activity shall be counted daily. S i j 3.8/4.8 170 i REV (11/15/73) 3 i
. .. -. . . _ - . . . - ~ . . - - - . . - . . _ . . . - . ~ ~ . . . _ _ . - . - . - . - - . - - - _ - -
a 1 l
. ~
.' ' 3.0 Lit 11TIriG C0 IDIT10;is or OPERATIOII 4.0 St3VEILLANCE REOUIREMEtJTS 4
.I
- a. Investigate to identify the 5. A determination shall be made of the
[ j causes for such release rates. total I-l'il released shall be performed on weekly. An analysis a sample at least
- b. Define and initiate a program monthly for I-133 and I-135.
to reduce such release rates 4 to the as low as practical levels. 6. A determination shall lie made of the j total radioactive particulates with half-
- c. Provide a report describing Ilves greater than 8 days released weekly .
these actions within 30 days as The particulate filters shall be removed an. unusual event (See Specifi- and analyzed for gross beta particulate , cation 6.7.B.2). radioactivity with half-lives greater j trian 8 days. Monthly, a composite of i 10. At least one of the . tw stack monitors, those filters used during the month
- including the charcoal cartridge and shall be prepared and analyzed for the
- particulate filter, shall be operable Principal gamma emitting radionuclides.
- at all times that the. stack is releasing j effluents to the environs. 7. Analysis for Sr-89 and Sr-90 shall be
] cade quarterly. Gross alpha radioactivity l 11. If both stack monitors are made or found in- shall be determined quarterly. j operable, the reactor shall be placed in the j hot standby condition within 24 hours. 12. Except as specified in 3.8.A.13, the off-gas stack and reactor building vent monitors shall have automatic isolation set points f consistent with Specification 3.8.A.1 and l' alarm set points consistent with Specifi-( ca tion. 3. 8.A.2. T l i
- 13. II operation is necessary with the of f gas
- j. Ilo1 dup System recombiners bypassed, the j' off gas stack monitors shall serve only
- an alarm function.
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3.0 LIMITING CONDITIONS FOR OPERA 110N 4.0 SURVEILIMCE REQUIREMENTS _
- 3. Two independent samples of each tank shall 3. The performance and results of independent be taken and analyzed for gross beta-gamma samples and valve checks shall be logged.
activity and the valve line-up checked prior to discharge of liquid effluents.
- 4. If the'11mits of 3.8.C cannot be met, radio- l active IIquid effluents shall not be released.
D. Radioactive Liquid Storage D. Radioactive Liquid Storage j The maximum gross radioactivity in liquid storage 1. A sample shall be taken, analyzed, and in-the Waste Sample, Floor Drain Sample, Waste recorded within 72 hours of each addition Surge, and Condensate Storage ~ Tanks shall be to a liquid waste storage tank to which
-less than 30 curies except for tritium and Specification 3.8.D. applies.
dissolved noble gases. If this condition
- cannot be met, the liquids in these tanks 2. If the sample analysis Indicates that the shall=be recycled to tanks.within the radwaste total radioactivity in the liquid waste L
facility until the condition is met storage tanks of Specification 3.8.D exceeds 30 curies, except for tritium and l dissolved noble gases, the liquids in these tanks shall be recycled to reduce the radioactivity to less than 30 curie; withia 24 hours of this saepling. E. Augmented ' Of f-Gas System E. Augmented Off-Gas System
- 1. If the hydrogen concentration in the of f- 1. The hydrogen monitors shall be functionally gas downstream of the recombiners reaches tested monthly and calibrated quarterly four percent, the recombiner off gas flow with an appropriate gas mixture source.
shall be stopped automaticr.lly by closing Each monitor shall have a sensor check the valves upstream of the recombiners. at least daily.
- 2. Except as specified in Specification i 2. Tank radiation monitors shall be calibraiad 3.8.E.3 below, at least one hydrogen quarterly by correlation with tank sampit monitor upstream and one hydrogen nwinitor analyses. Monitor readings shall be downstream of each operating recombinet recorded every eight hours to determine
. shall be operable during power operation. that the limit of Specification . 3.8.E.4 is not exceed 2d.
I 3.8/4.8 173 n-., ,,, ... O___ ._ m -
, - . _ . _____ _ - _ _ _ _ _ _ _ _ . . _ . _ . _ _ _ _ _ _ _ _ _ . _ . - _ - - - - . . - . - . _ _m.__. m_.__.._-- . , l t i l '3.0 LIMITING' CONDITIONS FOR OPER\ TION 4.0 SURVEILLVICE REQUIREMENTS t I ;
- 3. If the above specified upstream hydrogen 3. If a tank radiation monitor is t
monitors are not operable, continued inoperable, a sample from the gas operation of a-recombiner is permissible ! decay tank shall be taken, analyzed, i if the flydrogen Inventory Processor is ! and recorded every 24 hours. If no i set to provide a constant signal rep- additions to a tank have occurred
- l. resentative of the worse case hydrogen since the last sample, the tank need i concen t ra t ion. If the above specified not be sampled until the next additien.
- downstream hydrogen monitors are not
- operable, an orderly reactor shutdown ;
- shall be initiated to transfer the j i Off gas System to the recombiner bypass !
). mode. , i i , i 4. The maximum gross radioactivity contained ; l In one gas decay tank after 12 hours hold-l - up that can be discharged directly to the { environs shall be less than 22,000 curies ! of 'Xe-133 dose equivalent. .If these } conditions cannot. be net, the stored ! radioactive gas shall be recycled within
~
- 24 hours to other gas decay tanks until the condition is uet.
4- 5. During normal plant operation, radioactive l
, gaseous waste shall have'a minimum holdup i of 12 hours except for low radioactivity , ;. gaseous waste resulting from purge and ;
[ fill operations associated with refueling + 1 and reactor startup. Iloidup times for radio-active gaseous waste in the gas decay tanks l'
- j. shall be maximized consistent with plant
; . operation.
{ a 4 j F. Environmental Monitoring Program l 1 The environmental nonitoring program given j f u Table 4.8.1 shall be conducted. 1 l . i
~
a 3.8/4.g , 173A l uw (11/1;/71s 4
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t mh r oc so t_E oc jl o h b ad h ua m 4. t srd0 e rt og7b asi5 t u l 0h a ct ev i as o e rd 6 A sunewe o l i uu t eb cnn a uo h s t d cenrf obf t csh a saest dh udT go ce h ot ou oa ant e eia nt nl nd ammt i n nr ro , d s
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d a i u d t o . l a i euo r sreccit t2 t t osio0) pi t .l y ecenl ynnwml t sf ne snc s scee rivia gao eoe ea i S eaoat a aat v at aoc o ckR rb h em sb cere ai sh pi l e c mt t f srno rah wert di g ovaa ne f oeid ea Pt ov c nreso rI t O oh nadt h e u l i ee aetil o- , sG ct ot ed at ,hR o t sp t ih o eas S r sad en st F s saas agee ecsTr t vns eh i t b u nn C -e d a o eroaf h c eeoiaoom t il r arf e M eigot i shbt rg i o0t ot e maf . t " st - h eT c nt r1 u t sr , h oee naf 0 ew s asseoaf e 8 nsu nc e sa t me t f ol f 2 c ocuoc nn a eehil e . nsd t t uei( svm t nud eynrei t ii l eia uce nimn b co t s 5 el er s es ed ,bl s ur e i a niei r ah a n r otl oeia/eo uex eat sse 2 act P no dl eh rrceit l r a2 vyaanh oi urt ot if c f i m aoceot 5 semRi s eas ef nerie f u eqeo . f huol - nsoF t r vmek meosucs bl ed o 7 oerC ae i mrc r cece eh s e r et ihf rm t a aee h pl srt st t ,l z nt t 0 gb c gd t hh et) sa u ri i yamie oa c1 eu a l stTs g c odf om sgrul t il ne t s oeu onE si el ot i oemia t uosnn ih os g
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. umuo 'e o add sii ueiaod nd rr al n
dRl h nb aonck ed e h mbi aio . r ca eAf ese t o rb o s seb st o sr b) - eemic
- r l Sf h - s eie ce od d t er t m d guf i o iF eth n mth s e h a d ud oeAh esl e ramit b a t e at sl d t l l hl t msul0 oriexga nc r t eed uc eu ooe r ecutid b oso0
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s , a . B A Iflllj 'llllll,
._ _ -. .. . . . - _ - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . . . _ __.~m.m _ _ _ _ _ _ _ _ . .
4 [ d. 1 t i n - Detailed meteorological calculations for several locations off site have been made by the AEC staf f and the i most critical 22.5* sector was determ8ned to be at 600 m to the south-southeast at the site boundary. The j annual average dif fusion parameter va ue for the of f-gas stack release was determined to be 1.5 x 10-7 sec/m3 and for the reactor building vent release to be 7.2 x 10-6 sec/m3 l' i t The method utilized by the staf f to determine annual thyroid dose of 1500 mRen to a child for I-131 releases I j from the off gas stack and the reactor building vent is given in Fegulatory Guide 1.42. Based on this : method, the maximum I-131 concentration in milk from an existing cow would occur in the northwest sector at ! 1.5 miles which has an annual average dif fusion paraceter value of 1.7 x 10-8 sec/m3 for the of f gas stack-and 4.8 x 10-{ sec/ra3 for the reactor building vent. Based on these calculatio n , a continuous release rate of I-131 f rom the of f gas stack of 25 uCi/see or from the reac tor building vent of 0.9 uCi/sec could i resilt in an' annual thyroid dose of 1500 mrem to a child drinking this milk. i In order to limit I-131 releases in the gaseous effluents to as low as practical, quarterly average release j rates have been established which would require investigative actions at 2 percent of the maximum release ; rate and plant actions at-4 percent of-the maximum release rate. These release rates are significe tly below l
- 10 CFR Part 20 limits and'are factors of 2 and 4, respectively, above the as low as practical obje
- tIve cf l'i 1 percent of 10 CFR Part 20 limits.
i The AEC staf f perforced an analysis similar to that used to determine the maximum release rate of I-1 31 l i for the radioactive particulates with half-lives greater than 8 days. A reduction factor of 700 on the j MPCa to allow for possible ecological chain ef fects similar to those associated with the ccw-milk-child i thyroid for radioiodine was used. The annual average. dif fusion parameters at 600 m in the south-southeast I sector given previously wert used for both the off-gas stack and reactor building vent releases. Based en , these calculat. ions, a continuous release rate of ridioactive particulates with half-lives greater than 8 davs . I in the amount of 9.5 x 109 MPCa aci/sec from the off gas stack or 2 x 108 MPCa nCi/sec frem the reactor ; j building vent would not result in annual orga n doses in excess of the limits specified in 10 CFR Part 20.
- i In order to limit radioactive particulate releases in gaseous ef fluents t o as low as practical, quarterly
. average release rates have been established which would require investigative actions at 2 percent of the ; , maximum release rate'and plant actions at 8 percent of the maximum release rate. Tnese release rates are significantly below 10 CFR Part 20 limits and are factors of 2 and 8, respectively, above the as low as practical objectives of 1 percent of 10 CFR Part 20 limits. 4 t
! 177A i
- REV (11/13/73) l i <
1 n e n c y ny
* . l s o eg )
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- iwla tet t a 1 ec ee d gt ul aib r (
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- 8. heehvcmereai a cb o ,
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. g8 gi k5 maeor e r no f d priah c db ptI a3 ecb e hd oef o nmci e gssn 5 es t1 nt 9t i - sht oe t ecm snet o ut sd na ed a1 (nd t es eih cmesr ueigssn i sea c oid mneri eXgunt sat a r a nr - t eae sr S oeh us an esarp niefi r ,b6i e l
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i t aTl f k eooemo t s et m t cme .t v r e d alf c urcrii ne . ofd oo l n3 eei d rt eo ' nh e ot b d i sk roeHi ga mf f r g ao et f nel a sacf n t : rm r or r ;d oromr i f t ea s sf eer a r u i ga si .agem os uA s. n esh ot d t ssmMr nnagi f it s a d t u s awf nr f st er 1 asrsQ saee eeo ro nc e .er eg r eyee s na s r eoy d gvl sb a/ e . U a t t s o ien t nd pna uee sext J t eae nemg pl cmc n sac . hf t rod on o gl ar e n s ress t f c i er a yi - f n d ,inm a r siar enees b aph/ s g
. a o ja n eoaf t ri gtffi s eigge r eego m ooiiv s rif e erd n v t i e gs - r l c0 C sr arf i ei dhd snqenm d un e . eso no t pe E ea l i9h uao eeif v eood r c i p t gf nt al oe i e epd nnt eed ne woyi0 hl aon t e e a
m recred t v os cnr aoaoes oL pneh t l w0 en rt d s oeel ebi y l cc art soe 1 l o eee b ri . y e i st s bi f oejt h t rhl ari oS nf t c sr aoMgef , sia of ot h st me nt e ms wci asi sd t roesoan aeki ggnasd i r c es enr e y l nit miona r t amo aa eaC ne s aaxaerg ee ol pi pre ert es r n nnegt r sf s s : ge h a ,eiea cse anr ort ee aht eit u ua d yert avyt d e aS el i pl rt n m sas s eserl c ea raa s "s m .h e c pa oi es e al et eauh e e cneoemrt r iped ace s r eees ughh a eeich nd t r veeit h paa i s ei mr gi t a eiT ahd i t o g e gk snt t e rd o r e oserahd aa l er ee d s e t d s t l nc t ubi et eht oe h onrr nof a s ef roecia o pl r Mcat r Tf epoat oC h na sf osh ust seomef e T oguofI tNU s Ibt er op i~ l!! o
_ . - _ _ _ . _ . . _ _ ._ -_m ._ .m_. - t i
, 4
~
- i. Concentrations of gross radioactivity in the reactur building vent are expectest to be below the minimum l detectable levels with the existing analytical equipment. Therefore, isotopic analyses of samples from the ,
vent will not normally be performed. 5 Measuremen t of the gross radioactivity from the duct to the vent is based upon an equivaleat dose rate for the release rate in curies per second. Since an isotopic analysis canaat be made routinely of the vent ef fluent, the assumption is made that the isotopic corocsition in the vent vill be the sate as j
- determined at the steam jet ..ir ejector. Therefore, the average gamma energy per disintegration, E,y i i and the average beta energy disintegration, __E, to be used in the equation of Specification 3.8.A.1, will be based upon the average compositica of gases from the ai,r ejector unless the reactor building
} j vent sarple can be analyzed for its isatopic composition. The E shall3 be determined as previously j discussed for the off gas stack using the same reference data. The beta energy per disintegration fer j i those radiolsutupes determined to be present f rom the appropria te isotopic analysis shall be given in j USNRDL-TR- 802, II. Spectra of individual Negatron Emitters (Be ta Spect ra), II. Ilogan, P . E. Z i gean , and , 4 J. L. tbckin. The average beta energy shall be used as the beta energy per disintegration for each radio- i
- l. Isotope-evaluated. Using'these reference beta energies per disintegration with the appropriate composition ;
of radiogases in the vent, the average beta energy per disintegration, E r, will be determined. t The AEC staf f has performed an analysis to determine the equivalent dose rate (mR/hr) to the release rate given in Specification 3.8. A.1 if a typical of f gas mixture f rom the air e jector with 30 minutes delay and some f.uel failures is assumed. The relative E assured 7 was 0.7 Mev/ dis and the Eg was 0.3 Mev/ dis. ! i The resulting gamma dose rate for the radiation monitor equivalent to this release limit in the vent ; i (or uni t's duct) was determined to be,_3.3 mR/hr. Although only Ej is required for the determination of the ; release rate frora the of f gas stack, E3 is required to be applied for surveillance of releases from the i l i reactor building vent. b Determination of the I; and E gvalues to be us.-d in Epecification 3.8.A.I slull be performed weekly f rcm ! the appropriate isotople analysis until consistent values are obtained and quarterly thereaf ter unless l { changes are observed in either gaseous release _ rates of gross radioactivity or holdup time in the gas
, decay tanks. The quarterly determinations of Ey and Ea should be used in the eval,ation of compliance !
j with the quarterly release limits. The release of radioiodine from the off-gas stack and reactor building vent is moultored by the use of l charcoal cartridges which integrate the releases over the sarpling period of one to seven days. Frequency j of removal is dependent upon the release level measured on the previously rcraoved charcoal cartridge. . 4 The. analysis performed for 1-133 and I-135 Indicates the contribution of these radioiedines to the possible !
- =
inhalation doses.
~
178A j
; Rfv . e /isti n i
J 6 The release of radioactive particulates with half-lives greater than eight days from the off-gas stack and reactor building vent is monitored by the use of particulate filters which integrate the releases over the sampling period of one to seven days. All other aspects of particulate release measurements are similar to those discussed for radioiodine release measurements. The analysis performed for Sr-89 and Sr-90 and gross alpha radioactivity indicates the contribution of these radioisotopes to the gross particulate radioactivity. B. Mechanical vacuum Pump The purpose of isolating the mechanical vacuum pump line is to limit release of activity from the main condenser during a control rod drop accident. During the accident, fission products would be transported from the reactor through the main steamlines to the main condenser. The fission product radioactivity would i be sensed by the main steamline radioactivity monitors and initiate isolation. C. Liquid Effluents The radioactive liquid effluents from the Monticello plant will be. controlled on a batch basis with each l batch being processed by such method or methods appropriate for the quality of materials deternined to be present. Those batches in which the radioactivity concentrations are sufficiently low to allow release to the discharge canal are diluted with condenser circulating water in order to achieve the allowable con-centrations set forth in 10 CFR Part 20. The radioactive liquid w*ll be sampled and analyzed for gross radioactivity prior to release to the discharge canal, thus providing a means of obtaining information on effluents to be released so that appropriate release rates will be established. l Liquid effluent release will be controlled in terms of the concentrations in the discharge carut. In the case'of unidentified mixtures, such concentration limits are baaed on the assumption that the catire content is made up of the most restrictive isotope in accordance with 10 CFR Part 20. Such a limit assures that even if a person obtained all of his daily water intake from such a source, the resultant dose would not exceed that specified in 10 CFR Part 20. Since no such use of the discharge canal is made and considerable natural dilution occurs prior to any locations where such usage could occur, this assures that of fsite doses from this source will be far less than the limits specified in 10 CFR Part 20. If radioactive effluents are released to unrestricted areas on a radionuclide Lasis. the t:PC shall be determined and controlled in the cooling water discharge canal in accordance with Appendix B, Tabic II, Column 2 af'10 CFR Part 20 and Note I thereto. 179 REV (11/15/73)
i a l Detailed meteorological calculations for several locations of f site have been made by the AEC staf f and the most critical 22.5* sector was determined to be at 600 m to the south-southeast at the site boundary. The annual average diffusion parameter value for the off-gas stack release was determined to be 1.5 x 10-7 sec/m3 and for the reactor building vent release to be 7.2 x 10-6 sec/m3 The method utilized by the staff to determine annual thyroid dose of 1500 mrem to a child for I-131 releases from the of f-gas stack and the reactor building vent is given in Regulatory Guide 1.42. Based on this method, the maximum I-131 concentration in milk f rom an existing cow would occur in the northwest sector at 1.5 miles which has an annual average dif fusion parameter value of 1.7 x 10-8 sec/m3 for the of f-gas stack and 4.8 x 10-7 sec/m3 for the reactor building vent. Based on these calculations, a continuous release rate of I-131 from the off gas stack of 25 uCi/sec or from the reactor building vent of 0.9 uC1/sec could result in an annual thyroid dose of 1500 mrem to a child drinking this milk. In order to limit I-131 releases in the gasecus effluents to as low as practical, qturterly average release rates have been established which would require investigative actions at 2 percent of the maximum release rate and plant actions at 4 percent of the maximum release rate. These release rates are significantly below 10 CFR Part 20 limits and are factors of 2 and 4, respectively, above the as low as practical objective of 1 percent of 10 CFR Part 20 limits. The AEC staff performed an analysis similar to that used to determine the maximum release rate of I-131 for the radioactive particulates with half-lives greater than 8 days. A reduction factor of 700 on the MPCa to allow for possible ecological chain ef fects similar to those associated with the cow-milk-child thyroid for radioiodine was used. The annual average diffusion parameters at 600 m in the south-southeast 1 sector given previously were used for both the off gas stack and reactor building vent releases. Based on these calculations, a continuous release rate of radioactive particulates with half-lives greater than 8 days in the amount of . 9.5 x 109 MPCa uCi/sec from the of f-gas stack or 2 x 108 MPCa uCi/sec from the reactor building vent would not result in annual organ doses in excess of the limits specified in 10 CFR Part 20. In order to limit radioactive particulate releases in gaseous effluents to as low as practical, quarterly average release rates have been established which would require investigative actions at 2 percent of the maximum release rate and plant actions at 8 percent of the maximum release rate. These release rates are significantly below 10 CFR Part 20 limits and are factors of 2 and 8, respectively, above the as low as practical objectives of 1 percent of 10 CFR Part 20 limits. 177A REV (11/15/73) 4 4
s s The frequency for monitoring or sampling has been established so that if the euximum amount of gross radio-activity is exceeded, action can be taken to reduce the radioactivity to a level below the specified limit. F.- Environmental Monitoring Program it is recognized that a precise determination of environmental dose from a certain ec:ission from the stack is only possible by direct measurement. Such information vill be provided by the environmental monitoring program conducted at and around the site. If the stack emission ever reaches a level such that it is measureable in.the environment, such measurements will provide a basis for adjusting the proposed stack limit long-before the effect in the environment is of any concern for permissible dose. In this regard, it is ic:portant to realize that averaging emission rate over a period of one calendar year as permitted by 10 CFR Part 20. represents a very large safety turgin between conditions at any one instant (any minute, hour, or day) and the long-term dose of interest. 1 179B REV (11/15/73)
s T 4 s l (d) liighes t, lowest, and the annual average concentrations or levels of radiation for the sampling point with the highest average and description of the location of that point with respect to the site. (2) If levels of radioactive materials in environmental media as determined by an j environmental monitoring progran indicate the likelihood of public intakes in l excess of 1% of those that could result f rom continuous exposure to the l 4 concentration values listed in Appendix B, Table II, Part 20, estimates of the I likely resultant exposure to individuals and to population groups, and assumptions I upon which estimates are based shall be provided. (3) If statistically significant variation of offsite environmental concentrations with time are observed, correlation of these results with ef fluent release shall ' be provided.
- 1. Occupational Personnel Radiation Exposure (1) A tabulation of the number of occupational personnel exposures for. plant operations i personnel (permanent and temporary) in the following exposure increments for the !
reporting period: less than 100 mrem,100-250 mrem, 250-500 mrem, 500-750 cRem, 750-1000 mrem, 1-2 Rem, 2-3 Rem, 3-4 Rem, 4-5 Rem, 5-6 Ren, and greater than 6 Rem. ; (2) A tabulation' of the number of personnel receiving more than 500 mrem exposure in the reporting period according to duty function [e.g., routine plant surveillance and e inspection (regular duty), routine plant maintenance, special plant maintenance
, (describe maintenance), routine fueling operation, special refueling operation (describe '
operation), and other job-related exposures].
- l 1
(3) A tabulation annually of the number of personnel receiving more than 3 Rem and the }' major cause(s). l i 4 216 j 6.7 REV (11/15/73) ; a
I s B. Non-Routine Reports
- 1. Abnormal Occurrence Reports Notification shall be made within 24 hours by telephone and telegraph to the Director of the Regional Regulatory Operat ions Of fice (cc to the Director of Licensing), followed by a written report within 10 days to the Director of Licensing (cc to the Director of the Regional Regulatory Operations Office) in the event of the abnormal occurrences as defined in Section 1.0. The written report on these abnormal occurrences, and to the ext ent possible, the preliminary
' telephone and telegraph notification, shall: (a) describe, analyze and evaluate safety implications, (b) outline the measures taken to assure that the cause of the condition is dete rmined, (c) indicate the corrective action (including any changes made to the procedures and to the quality assur ice program) taken to prevent repetition of the occurrence and of similar occurrences involving similar components or systems, and (d) evaluate the saf ety implications of the incident in light of the cumulative experience obtained f rom the record of previous failures and malfunctions of similar systems and components.
6.7 ' 216A ret / (11/15/73)
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