ML19323H911
| ML19323H911 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/20/1980 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323H912 | List: |
| References | |
| NUDOCS 8006170249 | |
| Download: ML19323H911 (36) | |
Text
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NUCLE AR REGULATORY CON *, MISSION.
(G;j., M.*, s..
t V. ASHINGT ON. D. C. 20555 t,g~)4.4[j!
t WISCONS!!? ELECTRIC P0v:ER C0f'PANY DOCKET NO. 50-266 POINT EEACH f;UCLEAR PLANT, UNIT NO. 1 AfsEND"ENT TO FACILITY OPERATING LICEf;SE Anendment No.4 5 License No. DPR-24 i
1.
The Nuclear Regulatory Comnissicn (the Concission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated November 2, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will cperate in conforr.ity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorize:
by this anendnent can be conducted withcut endar.gering the heal.n and safety of the public, and (ii) that such activities will be conducted in conpliance with the Conmission's recc ations; i
D.
The issuance of this anendnent will not be inimical to the cara:n defense and security or to the health and safety of the public; and E.
The issuance of this amendnent is in accordance with 10 CFR Part 51 of the Comnission's reculations and all applicable requirere".s have been satisfied.
1 8006170 g
I I
2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating. License No. DPR-24 is hereby. amended to read as follows:
(B) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 45, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f
Y
.r.
Robert A. Clark, Chief Operating Reactors Branch # 3 Division of Licensing Attachnent:
Changes to the Technical Speci'fications Date of Issuance: May 20, 1980 4
+a
a
- TTACPMEf!T TG LICENSE Al EtsDMEt!TS A.E!;0MEf;T I;0. 45 TO FACILITY OPERATIf(G LICEf?SE f:0. DPR-24 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A as follows: -
Remove Pages Insert Pages 15-1 15-1 1
Table 15.3.5-2 Table 15.3.5-2 Table 15.3.5-2 (cont.)
Table 15.3.5-2 (cont.)
l Table 15.3.5-3 Table 15.3.5-3 Table 15.3.5-4 Table 15.3.5-4 15.3.15-1 15.3.15-2 15.3.15-3 Table 15.4.1-1 (cont.)
Table 15.4.1-1 (cont.)
Table 15.4.1-2 Table 15.4.1-2 Table 15.4.1-2 (cont.)
Table 15.4.1-2 (cont.)
Table 15.4.10-1 (cont.)
Table 15.4.10-1 (cont.)
15.4.15-2 15.4.15-2 15.6.9-10 15.6.9-10 15.6.11-1 15.6.11-1 i
1 I
i TABLE OF CONTENTS section Title Page 15 TECHNICAL SPECIFICATIONS AND BASES 15.1-1 15.1 Definitions 15.2.0 Safety Limits and Limiting Safety System settings 15.2.1-1 15.2.1-1 15.2.1 Safety Limit, Reactor Core 15.2.2-1
- 15.2.2 Safety Limit, Reactor Coolant System Pressure 15.2.3 Limiting Safety System Settings, Protective 15.2.3-1 Instrumentation 15.3-1 15.3 Limiting Conditions for Operation 15.3.1-1 15.3.1 Reactor Coolant System 15.3.2-1
- 15. 3.2 Chemical and volume Control System 15.3.3 Emergency Core Cooling System, Auxiliary Cooling 15.3.3-1 Systems, Air Recirculation Fan Coolers, and Containment Spray 15.3.4-1 15.3.4 Steam and Power Conversion System 15.3.5-1
- 15. 3.5 Instrumentation System 15.3.6-1 15.3.6 containment System 15.3.7-1 15.3.7 Auxiliary Electrical Systems 15.3.8-1 15.3.8 Refueling 15.3.9-1 15.3.9 Effluent Releases 15.3.10 Control Rod and Power Distribution Limits 15.3.10-1 15.3.11 Movable In-Core Instrumentation 15.3.11-1 15.3.12 Control Room Emergency Filtration 15.3.12-1 15.3.13 Shock Suppressors (Snubbers) 15.3.13-1 15.3.14-1,
15.3.14 Fire Protection System 15.3.15-1 15.3.15 Overpressure Mitigating System 15.4-1 15.4 Surveillance Requirements 15.4.1-1 15.4.1 Operational Safety Review 15.4.2 In-Service Inspection of Primary System Components 15.4.2-1 15.4.'3-1 15.4.3 Primary System Testing Following Opening 15.4.4-1 15.4.4 Containment Tests 15.4.5 Emergency Core Cooling System and Containment Cooling 15.4.5-1 System Tests 15.4.6-1 15.4.6 anergency Power System Periodic Tests 15.4.7-1 15.4.7 Main Steam Stop valves 15.4.8-1 15.4.8 Auxiliary Feedwater System 15.4.9-1 15.4.9 Reactivity Anomalies 15.4.10-1 15.4.10 Operational Environmental Monitoring 15.4.11-1 15.4.11 Control Room Emergency Filtration 15.4.12-1 15.4.12 Miscellaneous Radioactive Materials Sources 15.4.13 Shock Suppressors (Snubbers) 15.4.13-1 15.4.14-1 15.4.14 Surveillance of Auxiliary Building Crane 15.4.15-1 15.4.15 Fire Protection System 15-1
TABLE 15.3.5-2 I
1HSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP.
l 1
2 3
4 5
NO.OF MIN.
MINIMt.H PERMISSIBLE OPERATOR ACTIO NO. OF CHANNELS OPERABLE DEGREg OF BYPASS 1F CONDIT10 gig NO.
FUNCTIONAL UNIT CHANNELS TO CHANNELS REDUNDANCT CONDITIONS COLUtel 3 OR 4 TRIP CAltIOT BE BET :
1.
Manual 2
1 1
hintain hot shutdown 2.
Nuclear Flux Power Range **
2 di 4 power range Maintain hot low setting 4
2 3
2 channels greater than shutdown high setting 4
2 3
2 101 F.F. (low setting.only) 3.
Nuclear Flux Intermediate Range 2
1 1
2 of 4 power range channels Maintain hot greater than 10% F.P.
shutdown. Not 4.
Nuclear Flux Sourca Range 2'
1 1
1 of 2 intermediate range 10 channels greater than 10 shutdown. Not an9s 5.
Overtemperature AT 4
2 3
2 Maintain hot
'sh*utdown I
6.
Overpower AT 4'
2 3
2 h intain hot shutdown i
7.
Low Pressurizar Pressura 4
2 3
2 mintain het shutdown 8
Mi Pressuriser Pressure 3
~2 2
1 Maintain hot shutdown 9.
Pressurizer-Hi Water Level 3
2 2
1 Maintain hot shutdown
- 10. Low Flow in one loop (>50%
3/ loop 2/ loop 2
1 m intain hot F.P.)
(any loop) shutdown i
Low Flow Both Loops (10-60%
3/ loop 2/ loop F.F. ) -
(any loop)
Amendment No. 45
TABLE 15.3.5-2 (Cost'd) 1 2
3.
4 5
NO. OF MIN.
MINIMM FERMISSIBLE
, OPERATOR ACTICII NO. OF QlANNELS OPERAhlg nucarr OF SYPASS IF CONDITICIIS OF NO.
FUNCIIONAL IMIT Q!ANNEIS TO CRANNEI3 REDONDANCY CONDITIONS COLUMN 3 OR 4 TRIP CANNOT BE IET p.
- 11. Turbine Trip 3
2 2
1 Maintain <50I of rated power Maintain hot 12.
Steam Flow - Feed Water Flow 2floop 1/ loop 1/ loop 1/ loop shutdown
~
mismatch 13.
Io Lo Steam Generator 3floop 2/ loop 2/ loop 1/ loop liaintain het.
shutdown Water Level Maintain hot
- 14. Undervoltage 4 KV Bus 2/ bus 1/ bus 1/ bus (both buses) shutdown Maintain hot
- 15. Underfrequency 4 KY Bus 2/ bus 1/ bus 1/ bus shutdown (both buses)
Los individual rod 1
- 16. Control rod misaligament as 1
positions occa/ hour, acaitored by ee-line comyster and after a load change
>101 or aftse >30 inches of control red estion perg la men block eendicias exiaM. saintain aermal operettaa..
Full Fower F.F.
=
Not Applicable a
- One additional cha==at may be taken out of serv 4ce for zero power physics testing.
Amen / ment No. 45
TABLE 15.3.5-3 IMERGENCY COOI.ING 1
2 3
4 5
OPERAIOR AICTION NO. OF MIN.
MIN.
PERMISSIBLE IF CONDITIONS O NO. OF CHANNEL 's OPERABLE DEGREE OF BYPASS COEAnet 3 OR 4 NO.
FUNCTIGIAL UNIT OiANNELS TO TRIP QlANNELS REDUNDANCY OONDITIONS cam 4OT BE MET 1.
SAFETY IRLTECTION.
IIct Shutdown ***
a.
Manual 2
1 1
1 Hot Shutdownese b.
High Containment Pressure 3
2 2
1 Primary Pressure is steam [ Generator Iow Stesa Pressure /Imop 3
2 2
1 Less than 1000 psig Hot Shutdown ***
c.
d.
Pressurizer Iow Pressure 3
2 2
1 Primary Pressure is Less than 1800 peig Hot Shuteswn***
2.
00NTAINIENT SPRAY Not Shutdown **'
a..
Manual 2
2 2
l b.
Ri-Hi Oontainment Pressure 2 sets 2 of 3 2 per 1/ set Not Shutdown **8 (containment Spray) of 3 in each set set
- - laust actuate 2 switches siunaltaneously.
- - If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken on the affected unit t
to place the unit in cold shutdown conditions.
i s
Amendment No. 38 I
e
o TABLE 15.3.
4 INSTRIMENT OPERATING CONDITIONS FOR ISOLATI0tl FUNCTIONS 1
2 3
4 5
MINIMUN FERMISSIBLE
- 50. OF MIN.
DECREE BYFASS OPERATOR ACTIO5 NO. OF OlANNELS OPERABLE OF CONDITIONS IF CONDITIONS OF NO.
FUllCTIONAL UNIT CHANNELS TO CHANNELS REDU-COLT #983 OE 4 TRIP DANCT CANNOT RK MIT
- e I CONTAIIMENT ISOLATION l
a.
Safety Injection See Item No. I b,c, and d of Bot Shutdowname Table 15.3.5-3 b.
Manual 2
1 1
Bot Shutdown 2 STEAM LINE ISOLATION s.
El Ei Steam Flow with 2/Ioop 1
~
1 Safety Injection Ret Shutdown ***
b.
El Steam Flow 2/ loop I
1 Bot Shutdowm***
and 2 of 4 Low T with Safety Inje8EIon c.
Mi containment Pressure 3 2
2 1
Eet Shutdowm***
- j d.
Manual 1/1009 1/ loop 1/1**P not~ Shetdova M
- If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken os the affected unit to place
=
the unit in cold shutdown conditions.
Amendment No. 45 j
15.3.15 overpressure Mitigating System Op:;ratiens Applicability Applies to operability of the overpressure mitigating system when the reactor coolant system temperature is less than the minimuu temperature for the inservice pressure test.
Objective To s'pecify functional requirements and limiting conditions for operation on the.
use of the pressurizer power operated reifef valves when used as part of the overpressure mitigating system and to specify further limiting conditions for i
operation when the reactor coolant system is operated without a pressure absorbing volume in the pressurizer.
Specification A.
System Operability 1.
Except as specified in 15.3.15.A.2 below, the overpressurization miti-gating system shall be operable whenever the reactor coolant system is not open to the atmosphere and the temperature is less than the minimum pressurization temperature for the inservice pressure test, as specified in Figures 15.3.1-1 (Unit 1) and 15.3.1-3 (Unit 2). Operability require-i ments are:
Both pressurizer power operated relief valves operable at a setpoint a.
i of <1425 psig.
b.
The upstream isolation valves to both power operated relief valves are open.
2.
The requirements of 15.3.15.A.1 may be modified to allow one of the two power operated relief valves to be i'.op rabie for a period of not more than seven days.
knendment No. 45 15.3.15-1 s
-If the in:perable power operat:d reliof valves ccnnst be made cperable 3.
within seven days, the reactor coolant system must be depressurized and vented to the pressurizer relief tank within eight hours.
If both power operated relief valves are inoperable, the reactor coolant 4.
system must be depressurized and vented to the pressurizer relief tank within eight hours.
B.
Additional Limitations Een the reactor coolant system 1,s not open to the atmosphere and the
~
1.
tegerature of one or both reactor coolant system cold legs is <275'F, no'more than one high pressure safety injection pug shall be operable.
The second high pressure safety injection pump shall be demonstrated inoperable whenever the temperature of one or bbth reactor coolant system cold legs is <275'F by verifying that the motor circuit breakers have been removed from their electrical power supply circuits or by verifying that the discharge valves from the high pressure safety injection pumps to the reactor coolant system are shut and that power it removed from th* Terators.
A reactor coolant pump shall not be started when the reactor coolant 2.
system temperature is less than the minimum temperature for the inservice pressure test unless:
There is a pressure absorbing volume in the pressurizer or a.
The secondary water temperature of each steam generator is less than b.
50*F above the temperature of the reactor coolant system.
Basis The Overpressurization Mitigating System consists of a diverse means of relieving pressure during periods of water solid operation and when the system temperature i
This method below the value pemitted to perform the primary system leak test.
15.3.15-2 Amendment'No. 45
-rw-
of water relkf utilizes the pressurizer power operated relief valves (PORV's).
The PORV's are made operational for low pressure, relief by utilizing a dual set-point where the low pressure circuit is energized and de-energized by the operator with a keylock switch depending on plant conditions. The logic required for the l
low pressure setpoint is in addition to the existing PORY actuation logic and will l
l not interfere with existing automatic or manual actuation of the PORV's.
During plant cooldown prior to reducing reactor coolant system temperature below the minimum temperature allowable for the inservice pressure test, the operator unaer administrative procedures shall place the keylock switch in the " Low Pressure" position. This action enables the Overpressure Mitigating System. The redundant PORY channels shall remain enabled and operable while the. reactor coolant system is not open to the atmosphere and the temperature is less than the minimum pressuri-zation temperature for the inservice pressure test, except that one PORY may be out cf service for a period of up to seven days.
The mass input transient used to determine the PORV setpoint assumed a worse case transient of a single high pressure safety injection pump discharging to the reactor coolant system while the system is solid. Therefore, when the reactor coolant system is less than 275'F. only one high pressure safety injection pump shall be operable at any time except when the reactor coolant system is open to th'e atmosphere.
The heat input transient used to detennine the PORV setpoint assumes a temperature difference between the reactor coolant system and the steam generator of 50*F.
Therefore, before starting a reactor coolant pump when the reactor coolant system is solid, the operator shall insure that the secondary temperature of each steart, generator is less than 50*F above the temperature of the reactor coolant system unless a pressure absorbing volume has been verified to exist in the pressurizer.
knendment No. 45 15.3.15-3
~.
TABLE 15.4'.1-1 (CONTINUED)
Channel Descriptica Check Calibrate Test Remarks Harrow range co' tainment pressure -
^
24.
Contalammmat Pressure S
~R M**
n f
(-3.0, +3 p
- excluded) 7
- 25. ' steam' Generator Pressure S***
R M***
- 26. Turbine First Stage Pressure 38*
R M**
27.
Emergency Plan. Radiation M
R M
l Instruments 28 Envimatal Monitors M
N.A.
N.A.
l 29.
Overpressure Mitigating System 8
R Ms.nthl'y Each Shift M
S Prior to each startup if not done ' previous week Daily P
D Each Refueling Shutdown (But not to exceed 20 months, Weekly.
R w
~
except for first core cycle)
Biweekly a/w Mot applicable MA t
Not required during periods of refueling shutdown, but must be performed prior to starting up if it has not been performed during the previous surveillance period.
Not required during periods of refueling shutdown if steam generator vessel tesperature is greater than *10*F.
Each PORY ahall be demonstrated operable bys I
Performance of a channel functional test cn the PORY actuation channel, but excluding valve operation, a.
within 31 days prior to entering a condition in which the PORV is required operable and at least once per 31 days thereafter when the PORY is required operable.
l b.
Testing valve operation in accordance with the inservice test requirements of the ASME Boiler and Pressure Vessel Code,Section IX.
l Amendment No. A5
,l___
- _ s._
MINIMUM FREQUENCIES FOR EQUIPME:T AND SAMPLING TESTS Test Frequency 1.
Reactor Coolant Samples GrossBeta-gaEma 5/ week (7) activity (excluding tritium)
Tritium activity Monthly Radiochemical E Semiannually (2)
Determination (1)
Chloride concentration S/ week (8)
Diss. Oxygen Conc.
S/ week (6)
Fluoride Conc.
Weekly 2.
Reactor Coolant' Boron Boron Concentration Twice/ week 3.
Refueling Water Storage Boron Concentration Weekly (6)
Tank Watar Sample 4.
Boric Acid Tanks Boron Concentration Twice/ week 5.-
Spray Additive Tank NaOH Concentration Monthly 6..
Accumulator Boron concentration Monthly 7.
Spent Tual Pit Boron Concentration Monthly 8.
Secondary Coolant Gross Beta-gamma acti-Weekly (6) vity or gamma isotopic analysis Iodine concentration Weekly when gross Beta-gamma activity equals or exceeds 1.2 pci/cc (6) 9.
Control Rods Rod drop times of all Each refueling or full length rods (3) after maintenance that could affect proper func+4aning (4) 10.
Control Rod Partial movement of Every 2 weeks (6) all rods 11.
Pressurizer Safety Valver, Set point Each refueling shutdown 12.
P_ai ; Steam Safety Valves set point Each refueling shutdown 13.
Containment Isolation Trip Functioning Each refueling shutdown Amendnent No.-13,.2/6
g.
4 3 ___
Test _
Frecuency 14.
Refueling System Interlocks Functioning Each refueling shutdown 15.
Service Water System Functioning Each refueling shutdown 16.
Primary System Leakage Evaluate Monthly (6) 17.
Diesel Fuel Supply Fuel inventory Daily 18.-
Turbine Stop and Functioning Monthly (6)
Governor Valves 19.
Low Pressure Turbine Visual and magnetic Every five years Rotor Inspection (5) particle or liquid penetrant 20.
Boric Acid System Storage Tank Daily Temperature 21.
Boric Acid System Visual observation Daily of piping temperatures (all ll45'F) 22.
Boric Acid Piping Heat Electrical circuit Monthly Tracing operability (1)
A radiochemical analysis for this purpose shall consist of a quantitative measure-ment of each radionuclide with half life of >30 minutes such that at least 95%
o_f total activity of primary coolant is accounted for.
(2)
E determination will be started when the gross activity analysis of a filtered sample indicates 110 pc/cc and will be redetermined if the primary coolant gross radioactivity of a filtered sample increases by more than 10 pc/cc.
(3)
Drop tests shall be conducted at rated reactor coolant flow.
Rods shall be dropped under both cold and hot conditions, but cold drop tests need not be timed.
(4)
Drop tests will be conducted in the hot condition for rods on which maintenance was performed.
(5)
As accessible without disassembly of rotor.
(6)
Not required during periods of refueling shutdown.
(7)
At least once per week during periods of refueling shutdown.
(8)
At least three times per week (with maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples) during periods of refueling shutdown.
/
Amendment No. 1, 12, 27, 32, 45
TABLE 15.4.10-1 (CONTINUED)
Sample Type Incations (a,c)
Frequency' Analysis Commento Woll Water 1-Onsite Well (10)
Quarterly Gross Deta-T.S.(b)
Gausna Scan T.S.
Tritium Strontium-89 Strontium-90 Milk 1-Incal dairy pool (11)
Monthly Gamma Scan Radioiodine analysis 1-Dairy Farm, NNW (19)
Radioiodine done by the resin 1-Dairy Farm, SSE (21)
Strontium-89 extraction technique.
Strontium-99 Algae 1-North of Discharge (5) 3x/yr Gross Beta 1-Discharge of Flume (12) as available Ganaa Scan Fish 1-Travelling screens (13) 3x/yr Gross Beta Analysis of edible as available Gamma Scan portions only.
(a)
Reference location is chosen well in excess of 10 miles from the plant in a low X/Q sector to provide an estimate of background levels.
(b)
T.S. - Total Solids (c)
Numbers given under location correspond to sampling locations shown in Figure 15.4.10-1.
l 1
Amendment No. 45 Page 2 of 2 l
m
1 C.
Fira Rosa Sr.ations Test Froggety 1.
Visual Inspection Monthly 2.
Rose Hydro-Test Yearly 3.
Partially open each hose station valve 3 years j
to verify operability and no haeckage D.
Fire Detection Test Fregtpemey 1.
Channel Functional Test 2 30.
E.
Fire Barrier Penetration Fire Seals Tost Frostuency 1.
Visual Inspection is me. and following repairs or seeintenance F.
Fire Fanp Diesel Engine Test p cy 1.
a.
Varify 200 galloes of feel in Meathly fuel storage tank b.
Verify diesel starts from ambient Monthly conditions and operates for at least 20 minutes.
2.
Sample diesel fuel per AS1M-D270 Qir. ;terly and verify acceptable per Table 1 of AS1H-D975 with respect te viseesity, water content and sadimt.
3.
a.
Inspect diesel per preseenres 18 wh prepared in conjunction with its manufacturer's recommendations b.
Verify diesel starts frem W ent 18 months conditions and operates for 120 minutes while loaded with the fire pung Amendment No. 32, 45 15.4.15-2
(1) Tne n==ber and types of samples taken and the measure-
=ents _ade on the samples; e.g., gross beta ga=na scan, etc.
(2) Any changes _.ade in sample types or locations during the reperting period, and criteria for these changes, b.
A su==ary of survey results during the reporting period.
4.
Leak Testing of Source Results of required leak tests performed on seal sources if the tests reveal the presence of 0.005 microcuries or more of re=ovable conta=ination.
D.
Poison Assembly Re= oval from Spent Fuel Storage Racks Plans for removal of any poison assemblies from the spent fuel storage racks shall be reported and described at least 14 days prior to the planned activity.
Such report shall describe neutron attenuation testing for any replacement poison assemblies, if applicable, to confirm the presence of boron =aterial.
E.
overoressure Mitigating System operation In the event the overpressure mitigating system is operated to relieve a pressure transient which, by licensee's evaluation, could have resulted in an overpressurization incident had the system not been operable; a special report shall be prepared and submitted to the Co: mission within 30 days. The report shall describe the circumstances initiating the transient, the effect of the system on the transient and any corrective action necessary to prevent recurrence.
Amend:ent No. 35, 45 15.6.c-10
15.6.11.
RADIATION PROTECTION PROGRAM specification Radiological control procedures shall be written and made available to all station The radiation perconnel, and shall state permissible radiation exposure levels.
Protection program shall meet the requirements of 10 CFR 20, with the exception of tho followings Paragraph 20.203 - Caution signs, labels and signals In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c) (2), each radiation area in which the intensity of radiation is greater than 100 mrem /hr shall be barricaded and conspicuously posted as a High Radiation Arca, and entrance thereto shall be controlled in accordance with the Point Beach Nuclear Plant Health Physics Mm4Mstrative Control Policies and Procedure Manual, S:ction 2.7, Radiation Work Permit. A person or persons permitted to enter such arcas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.
In addition, each High Radiation Arca outside the containment building in which the intensity of radiation is greater then 1000 mrem /hr shall be prov$ded with locked barricades to prevent unauthorized entry into these areas, and the keys to these locked barricades shall be maintained under the ad=inistrative control of the Duty Shift Supervisor.
Amendment tio. JB 45 15.6.11-1 l
t
- . 8 4
- .,'%
UNITED STATES f'
NUCLE AR REGULATORY COMMISSION f
- y.r. ff 7 wAsmucTou. o. c. 20sss s,.
g.
%,l+; $l 1.'ISCONSIN ELECTRIC POWER C0i?A!:Y DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT NO. 2 A!'.E::D!'.ENT TO FACILITY OPERATING LICENSE Anendment No. 50 License No. DPR-27 1.
The Nuclear Regulatory Conmission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated November 2,1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Conmission; C.
There is reasonable assurance (i) that the activities authorized by this amerdment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and securi. y or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable recuirements have been satisfied.
N1 vd 4 2-Accordingly, the license is acended by changes to the Technical
'?
Specifications as indicated in the attachnent to this license anedent, and paragraph 3.B of racility Operatin; Licer.se No. DPR-27 is' hereby,anended,to read as follows:
(B) Technical Specifications The Technical' Specifications contained in Appendices A and B, as revised throt.gh Amendment No. 50, are The licensee shall hereby incorporated in the license.
operate the facility in accordance with the Technical Specifications.
3.
This license ' amendment is effective as' of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISS:0N 0 -p. f. [f t _ rsku g
obert A. C' ark, ie ~
Operating Reactors Branch ! 3 Division of Licensing
- tach.ent:
Changes to the Technical Specifications Cate of Issuance: May 20. 1980 4
ATTACHMENT TG LICENSE AMENDMENTS A".ENDMENT N0. 50 TO FACILITY OPERATING LICENSE NO. D?R-27 DOCKEf NOS. 50-266 AND 50 -301 Re.ise Appendix A as follows:
Remove Pages Insert Pages 15-1 15-1 Table 15.3.5-2 Table 15.3.5-2 Table 15.3.5-2 (cont.)
Table 15.3.5-2 (cont.)
Table 15.3.5-3 Table 15.3.5-3 Table 15.3.5-4 Table 15.3.5-4 15.3.15-1 15.3.15-2 15.3.15-3 Table 15.4.1-1 (cont.)
Table 15.4.1-1 (cont.)
Table 15.4.1-2 Table 15.4.1-2 Table 15.4.1-2 (cont.)
Table 15.4.1-2 (cont.)
Table 15.4.101 (cont.)
Table 15.4.10-1 (cont.)
15.4.15-2 15.4.15-2 15.6.9-10 15.6.9-10 15.6.11-1 15.6.11-1
TABLE OF CONTENTS Section Title Page 15 TECHNICAL SPECIFICATIONS AND BASES 15.1-1 15.1 Definit 1ons 15.2.0 Safety Liaits and Limiting Safety System SettingL 15.2.1-1 15.2.3-1 15.2.1 Safety Limit, Reactor Core 15.2.2 Safety Limit, Reactor Coolant System Pressure 15.2.2-1 15.2.3 Limiting Safety System Settings, Protective 15.2.3-1 Instrumentation 15.3-1 15.3 Limiting Conditions for Operation 15.3.1-1 15.3.1 Reactor Coolant System 15.3.2 Chemical and Volume Control System 15.3.2-1 15.3.3 Emergency Core Cooling System, Auxiliary Cooling 15.3.3-1 Systems, Air Recirculation Fan Cooler,s, and Containment Spray 15.3.4 Steam and Power Conversion System 15.3.4-1 15.3.5 Instrumentation System 15.3.5-1 15.3.6-1 15.3.6 Containment System 15.3.7 Auxiliary Electrical Systems 15.3.7-1 15.3.8-1 15.3.8 Refueling 15.3.9 Effluent Releases 15.3.9-1 15.3.10 Control Rod and Power Distribution Limits 15.3.10-1 15.3.11 Movable In-Core Instrumentation 15.3.11-1 15.3.12 Control Room Emergency Filtration 15.3.12-1 15.3.13 Shock Suppressors (Snubbers) 15.3.13-1 15.3.14 Fire Protection System 15.3.14-1 15.3.15 Overpressure Mitigating System 15.3.15-1 15.4 Surveillance Requirements 15.4-1 15.4.1-1 15.4.1 Operational Safety Review 15.4.2 In-Service Inspection of Primary System Components 15.4.2-1 15.4.3 Primary System Testing Following Opening 15.4.3-1 15.4.4 Containment Tests 15.4.4-1 15.4.5 Emergency Core Cooling System and Containment Cooling 15.4.5-1 System Tests 15.4.6 Emergency Power System Periodic Tests 15.4.6-1 15.4.7 Main Steam Stop Valves 15.4.7-1 15.4.8-1 15.4.8 Auxiliary Feedwater System 15.4.9-1 15.4.9 Reactivity Anomalies 15.4.10-1 15.4.10 Operational Environmental Monitoring 15.4.11 Control Room Emergency Filtration 15.4.11-1 15.4.12 Miscellaneou's Radioactive Materials Sources 15.4.12-1 15.4.13 Shock Suppressors (Snubbers) 15.4.13-1 15.4.14 Surveillance of Auxiliary Buildir.9 Crane 15.4.14-1 15.4.15-1 15.4.15 Fire Protection System A=endment No. 16, 26, 16, 5 0 15-i L
TABLE 15.3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP.
1 2
3 4
5 NO.OF MIN.
MINIMUM PERMISSABLE OPERATOR ACTION NO. OF CHANNELS OPERABLE DEGREE OF BYPASS IF CONDITIONS OF NO.
FUNCTIONAL UNIT CHANNELS TO CHANNELS REDUNDANCT CONDITIONS COLUMN 3 OR 4 TRIP CANNOT BE MET i
1.
Manual 2
1 1
hintain hot shutdown 2.
Nuclear Flux Power Range **
2 of 4 power range Maintain hot low setting 4
2 3
2 channels greater than shutdown high setting 4
2 3
2 10% F.P. (low setting.only) 3.
Nu.: lear Flux Intermediate Range 2
1 1
2 of 4 power range channels hintain hot greater than 10% F.P.
shutdown. Note 1 4.
Nuclear Flux Source Range 2'
1 1
1 of 2 intermediate range h intain hot gg channela greater than 10 shutdown. Note 1 amps 5.
Overtemperature at 4
2 3
2 Maintain hot shutdown 6.
Overpower AT 4
2 3
2 hintain hot shutdown
~
- 7., Low Pressurizer Preasura 4
2 3
2 hintain het shutdown 8.
Hi Pressuriser Prassure 3
2 2
1 Maintain hot shutdown 9.
Pressurizer-Hi Water Level 3
2 2
1 Maintain hot shutdown
- 10. Low Flow in one loop (>50%
3/ loop 2/ loop 2
1 Haintain hot F.P.)
(any loop) shutdown Low Flow Both Loops (10-5'0%
3/ loop 2/ loop F.P. )
(any loop)
Amendment No. 5 0
TABLE 15.3.5-2 (Cont'd)
~
1 2
3 4
5 NO. OF MIN.
MINIMUM PERMISSIBLE OPERATOR ACTION NO. OF OIANNELS OPERABLE DECREE OF BYPASS IF (DNDITIONS OF NO.
FUNCTIONAL UNIT OIANNELS TO OIANNEIS REDUNDANCT CONDITIONS COLUMN 3 OR 4 TRIP CANNOT BE NET
- 11. Turbine Trip 3
2 2
1 Maintain <50% of rated Power
- 12. Steam Flow - Feed Water Flow 2/ loop 1/ loop 1/ loop 1/ loop Maintain hot
~
shutdown mismatch
- 13. Io to Steam Cenerator 3/ loop 2/ loop 2/ loop 1/ loop M'aintain hot shutdown Water Level Maintain hot
.14.
Undervoltage 4 KV Bus 2/ bus 1/ bus 1/ bus shutdown j
(both buses)
Maintain hot
- 15. Underfrequency 4 KV Bus 2/ bus 1/ bus 1/ bus shutdown (both busas)
Log individual rod 1
- 16. Control rod misalignment as 1
positions once/ hour, monitored by on-line computer and after a load change
>10% or ef ter >30 inches of contrci rod motion When block condition exists, maintain normal operation.
HOTE 1:
Full Power F.F.
=
l O Not Applicable do One additional channel may be taken out of service for zero power physics testing.
i Amendment No. 5 0
TABLE 15.3.5-3 EMERGENCY COOLING 1
2 3
4 5
OPERATOR ACTION NO. OF
- MIN, MIN.
PERMISSIBLE IF CONDITIONS W NO. OF CHANNEIS OPERABIJE DEGREE OF BYPASS COLUMN 3 OR 4 NO.
FUNCTIONAL IMIT CHANNELS TO TRIP CHANNELS REDUNDANCY CONDITIONS CANNOT BE MET 1.
SAFETY INJECTION Hot Shutdown ***
o.
. Manual 2
1 1
1 Hot Shutdown ***
b.
High Containment Pressure 3
2 2
1 Primary Pressure is Steam Generator Iow Steam Pressure /Icop 3
2 2
1 Iess than 1800 psig Hot Shutdown ***
c.
d.
Pressurizer Low Pressure 3
2 2
1 Primary Pressure is Less than 1800 psig Hot Shutdown **4 2.
CONTAINMENT SPRAY Mot Shutdown ***
2 2
2 a.
Manual b.
Hi-Hi Containment Pressure 2 sets 2 of 3 2 per 1/ set Hot Shutdown ***
(Containment Spray) of 3 in each set set i
- - Mus,t actuaEs 2 switches simult aneously.
- - If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken on the affected unit J
j to place the unit in cold shutdown conditions.
1 l
Amendment No. n,5 0 l
TABLE 15. 3. 5-4
~
INSTRUMENT OPERATING CONDITIONS FOR ISOLATION PUNCTIONS 1
2 3
4 5
MINIMUM PERMISSABLE NO. OF MIN.
DEGREE BYPASS OPERATOR ACTI0tt '
NO. OF CHANNELS OPERABLE OF CONDITIONS IF CONDITIONS OF NO.
FUNCTIONAL UNIT CHANNELS TO CHANNELS REDU-COLUMN 3 OR 4 TRIP DANCY CANNOT BE MET 1 CONTAINMENT IS0!ATION o.
Safety Injection See Item No. 1 b,c, and d of
~
Hot Shutdown ***
Table 15.3.5-3 b.
Manual 2
1 1
Hot Shutdown 2 STEAM LINE IS0lATION HiHiSteamFkowwith 2/ loop 1
1 a.
Safety Injection Not Shutdown ***
b.
Hi Steam Flow 2/ loop 1
1 Ho t Shu td own* **
and 2 of 4 Low T with Safety Inje$lfon c.
Mi Containment Pressure 3 2
2 1
Bet Shutdown ***
d.
Manual 1/l**p 1/l**P 1/100P Not Shutdown i
i eco - If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken on the affected unit to place the unit in cold shutdown conditions.
4 i
Amtndment No. 5 0
15.3.15 Overpressure Mitigating System Operations
_ Applicability
' Applies to operability of the overpressure mitigating system when the reactor coolant system temperature is less than the minimum temperature for the inservice pressure test.
Objective To specify functional requirements and limiting conditions for operation on the use of the pressurizer power operated relief valves when used as part of the overpressure mitigating system and to specify further limiting conditions for operation when the reactor coolant system is operated without a pressure absorbing volume in the pressurizer.
Specification A.
System Operability 1.
Except as specified in 15.3.15.A.2 below, the overpressurization miti-gating system shall be operable whenever the reactor coolant system is not open to the atmosphere and the temperature is less than the minimum pressurization temperature for the inservice pressure test, as specified in Figures 15.3.1-1 (Unit 1) and 15.3.1-3 (Unit 2). Operability require-ments are:
a.
Both pressurizer power operated relief valves operable at a setpoint of <425 psig.
i b.
The upstream isolation valves to both power operated relief valves l
are open.
2.
The requirements of 15.3.15.A.1 may be modified to allow one of the two power operated relief valves to be inoperable for a. period of not more than seven days.
A=a: ent No. 5 0 15.3.15-1
s 3.
.If the inoperable power operated relief valves cannot be made operable
)
within seven days, the reactor coolant system must be depressurized and f
Vented to the pressurizer relief tank within eight hours, i
4.
If both power operated relief valves are inoperable, the reactor coolant system must be depressurized and vented to the pressurizer relief tank within.eight hours.
B.
Additional Limitations 1.
When the reactor coolant system is not open to the atmosphere and the tagerature of.one or both reactor coolant system cold legs is <275'F, no mon than one high pressure safety injection pug shall be operable.
The second high pressure safety injection pump shall be demonstrated i
inoperable whenever the temperature of one or both reactor coolant system cold legs is <275'F by verifying that the motor circuit breakers j
l have been removed from their electrical power supply circuits or by verifying that the discharge valves from the high pressure safety injection pumps to the reactor coolant system are shut and that power is removed from their operators.
j 2.
A reactor coolant pug shall not be started when the reactor coolant system temperature is less than the minimum temperature for the inservice j
pressure test unless:
There is a pressure absorbing volume in the pressurizer or a.
b.
The secondary water temperature of each steam generator is less than 1
50*F above the temperature of the reactor coolant system.
Basis-The Overpressurization Mitigating System consists of a diverse means of relieving f
pressure during periods of water solid operation and when the system temperature is j
below the.value permitted to perfonn the primary system leak test. This method t
l
[
Amendmentf No. 5 0 15.3.15-2 l
of water relief utilizes the pressurizer power operated relief valves (PORV's).
The PORV's are made operational for low pressure relief by utilizing a dual set-point where the low pressure circuit is energized and de-energized by the operator with a keylock switch depending on plant conditions. The logic required for the low pressure setpoint is in addition to the existing PORV actuation logic and will not interfere with existing automatic or manual actuation of the PORV's.
During plant cooldown prior to reducing reactor coolant system temperature below the minimum temperature allowable for the inservice pressure test, the operator unaer administrative procedures shall place the keylock switch in the " Low Pressure" position. This action enables the Overpressure Mitigating System.
The redundant PORY channels shall remain enabled and operable while the reactor coolant system is not open to the atmosphere and the temperature is less than the minimum pressuri-zation temperature for the inservice pressure test, except that one PORY may be out of service for a period of up to seven days.
The nass input transient used to determine the PORV setpoint assumed a worse case transient of a single high pressure safety injection pump discharging to the reactor coolant system while the system is solid. Therefore, when the reactor coolant system is less than 275'F only one high pressure safety injection pump shall be operable at any time except when the reactor coolant system is open to the atmosphere.
The heat input transient used to determine the PORV setpoint assumes a temperature diffemca between the reactor coolant system and the steam generator of 50*F.
Therefore, before starting a reactor coolant pump when the reactor coolant system is solid, the operator shall insure that the secondary temperature of each steam generator is less than 50*F above the tettperature of the reactor coolant system unless a pressure absorbing volume has been verified to exist in the pressurizer, t
l Acend:ent No. 5 0 15.3.15-3
~
TABIJ 15.4,1-1 (CCm inusu)
Channel' Description Check Calibrate Test Remarks
- 24.. Containment Pressure S
R M**
Harrow range containment-pressure
(-3.0, +3 psig excluded) 25.
Steam Generator Pressure S***
R M***
26.
Turbine First Stage Pressure S**
R M**
- 27.. Emergency Plan Radiation "4
R M
Instruments 28 Environmental Monitors M
N.A.
N.A.
29.
Overpressure Mitigating System 8
R l
Monthly Each Shift M
S Prior to each startup if not done previous week Daily P
D Each Refueling Shutdown (But not to exceed 20 months, Weekly R
W except for first core cycle)
Biweekly B/W Not applicable NA l
. Not required during periods of refueling shutdown, but must be performed prior to starting up if it has not been performed during the previous surveillance period.
Not required during periods of refueling shutdown if steam generator vessel temperature is greater than 70*F.
Performance of a channel functional test on the PORV actuation channel, but excluding valve operation, a.
within 31 days prior to entering a condition in which the PORV is required operable and at least once per 31 days thereafter when the PORV is required operable.
b.
Testing valve operation in accordance with the inservice test requirements of the ASME Boiler and Pressure Vessel Code,Section IX.
0 Amendment No.
4
--TABLE 13gCs1 ~
~
MINIMUM FREQUENCIES FOR EQUIPMENT AND SAMPLING TESTS l
Test Frequency 1.
Reactor Coolant Samples Gross Beta-gamma S/ week (7) activity (excluding tritium)
)
Tritium activity Monthly Radiochemical E Samiannually (2)
Dett.rmination (1) f i
Chloride Concentration 5/ week (S) l 1
I Diss. Oxygen Conc.
5/ week (6)
Fluoride Conc.
Week?.y 2.
Reactor Coolant Boron Boron Concentration Twice/ week 3.
Refueling Water Storage Boron Concentration Weekly (6)
Tank Water Sample 4.
Boric Acid Tanks Boron Concentration Twice/ week 5.
Spray Additive Tank NaOH Concentration Monthly 6.
Accel ator Boron Concentration Monthly 7.
Spect Tual Pit Boron Concentration Monthly 8.
Secondary Coolant Gross Beta-gassa acti-Weekly (6) vity or gasma isotopic analysis Iodine concentration Weekly when gross meta-gamuna activity equals or exceeds 1.2 pCi/cc (6) 9.
Control Rods Rod drop times of all Each refueling or full length rods (3) after maintenance that could affect proper functioning (4) 10.
Control Rod Partial movement of Every 2 weeks (6) all rods 11.
Pressurizar Safety Valves Set point Each refueling shutdown 12.
Main Steam Safety Valves set point Each refueling shutdown
.13.
Containment Isolation Trip Functioning Bach refueling shutdown i
Amtnd ent :;o. 17, 31, 0
TABLE 15.4.1-2 (CONTINUED)
Test Frequency 14.
Refueling System Interlocks Functioning Each refueling shutdown 15.
Service Water System Functioning Each refueling shutdown 16.
Primary System Leakage Evaluate Monthly (6) 17.
Diesel Fuel Supply Fuel inventory Daily 18.
Turbine Stop and Functioning Monthly (6)
Governor Valves 19.
Low Pressure Turbine Visual and magnetic Every five years Rotor Inspection (5) particle or liquid penetrant 20.
Boric Acid System Storage Tank Daily Temperature 21.
Boric, Acid System Visual observation Daily of piping temperatures (all >145'F) 22.
Boric Acid Piping Heat Electrical circuit Monthly Tracing operability (1)
A radiochemical analysis for this purpose shall consist of a quantitative measure-ment of each radionuclide with half life of >30 minutes such that at least 95t o_f total activity of primary coolant is accounted for.
l (2)
E detemination will be started when the gross activity analysis of a filtered sample indicates >10 pc/cc and will be redetermined if the primary coolant gross radioactivity of a filtered sample increases by more than 10 pc/cc.
(3)
Drop tests shall be conducted at rated reactor coolant flow.
Rods shall be dropped under both cold and hot conditions, but cold drop tests need not be timed.
(4)
Drop tests will be conducted in the hot condition for rods oa which maintenance was performed.
(5)
As accessible without disassembly of rotor.
(6)
Not required during periods of refueling shutdown.
l (7)
At least once per week during periods of refueling shutdown.
l (8)
At least three times per week (with maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples) during periods of refueling shutdown.
Amendment No.38 5 0
,-.w-,
TABLE 15.4.10-1 (CONTINUED)
Sampie Type Iocations(*is)
Frequency Analysis Cosaments Wall Water 1-Onsite Well (10)
Quarterly Gross Beta-T.S.(b)
Gamuna Scan T.S.
Tritium Strontissa-89 Strontium-90 Milk 1-Iocal dairy pool (11)
Nonthly Ga m Scan Radioiodine analysis 1-Dairy Farm, NNW (19)
Radioiodine done by the resin 1-Dairy Farm, SSE (21)
Strontitua-89 extraction technique.
Strontium-9p Algae 1-North of Discharge (5) 3x/yr Gross Beta 1-Discharge of Flume (12) as available Gasona Scan i
Fish 1-Travelling screens (13) 3x/yr Gross Beta Analysis of edible as available Gamma Scan portions only.
I i
j (a)
Reference location is chosen well in excess of 10 milec from the plant in a low X/Q sector to provide an estimate of background levels.
(b)
T.S. - Total Solids (c)
Numbers given under location correspond to sampling locations shown I
in Figure 15.4.10-1.
I
}
i 1
4 Amendment No. 25 Page 2 of 2
l C.
Fira Mos3 station]
Test Frequency 1.
Visual Inspection Monthly 2.
Rose Hydro-Test Yearly 3.
Partially open each hose station valve 3 years to verify operability and no blockage D.
Fire Detection Test Frequency 1.
Channel Functional Test 2 mo.
E.
Fire Barrier Penetration Fire Seals Test Frequency 1.
Visual Inspection 18 mo. and following repairs or maintenance F.
Fire Fump Diesel Engine Test Frequency 1.
a.
Varify 200 gallons of fuel in Monthly fuel storage tank b.
Verify diesel starts from ambient Monthly conditions and operates for at least 20 minutes.
1 2.
Sample diesel fuel per ASTM-D270 Quarterly and verify acceptable per Table 1 of AS'DI-6975 with respect to viscosity, water content and sediment.
3.
a.
Inspect diesel per procedures 18 months prepared in conjunction with its manufacturer's reconumendations b.
Verify diesel starts from ambient 18 months conditions ind operates for >20 minutes while loaded with the fire pump i
Arendment No. 36 15.4.15-2
(1) The ne=ber and typas of sa=ples teken and tha measure-
=ents =ade on the samples ;
e.g., gross beta gamma scan, etc.
(2) Any changes made in sample types or locations during the reporting period, and criteria for these changes.
b.
A su==ary of survey results during the reporting period.
4.
Leak Testing of Source Results of required leak tests performed on seal sources if the tests reveal the presence of 0.005 microcuries or more of removable conta=ination.
D.
Poison Assembly Renoval from Spent Fuel Storage Racks Plans for removal of any poison assemblies from the spent fuel storage racks shs11 be reported and described at least 14 days prior to the planned activity.
Such report shall describe neutron attenuation testing for any replacement poison assemblies, if applicable, to confirm the presence of boron =aterial.
E.
oveirpressure Mitigating system operation In the event the overpressure mitigating system is operated to relieve a prese_. e transient which, by licensee's evaluation, could have i
resulted in an overpressurization incident had the system not been operable a special report shall be prepared and submitted to the Consnission within 30 days. The report shall describe the circumstances initiating the transient, the effect of the system on the transient and any corrective action necessary to prevent recurrence.
Amendment No. 41 15.6.9-10 l
l l
15.6.11 RADIATION PROTECTION PROGRAM Specification Radiological control procedures shall be written and made availtble to all station personnel, and shall state permissible radiation exposure levels. The radiation protection program shall meet the requirements of 10 CFR 20, with the exception of the following:
Paragraph 20.203 - Caution signs, labels and signals In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c) (2), each radiation area in which the intensity of radiation is greater than 100 mrea/hr shall be barricaded and conspicuously posted as a High Radiation i
Area, and entrance thereto shall be controlled in accordance with the Point Beach Nuclear Plant Health Physics Administrative Control Policies and Procedure Manual, Section 2.7, Radiation Work Permit. A person or persons permitted to enter such arear shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area. In addition, each High Radiation Area outside the containment building in which the intensity of radiation is greater than 1000 mrom/hr shall be provided with locked barricades to prevent unauthorized entry into these areas, and the keys to these locked barricades shall be maintained under the administrative control of the Duty Shift Supervisor.
Amendment No. ' u, 5 0 15.6.11-1
_ -,. _. _. _. _.-