Text

Proposed Tech Specs Re Rev of Statement on Normal Shift Length & Other Misc Changes
	ML20154L955
Person / Time
Site:	Monticello
Issue date:	10/12/1998
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20154L936	List: ML20154L932; ML20154L940; ML20154L955;
References
	NUDOCS 9810200125
	Download: ML20154L955 (98)
	v • d • e

...- -.

. _ = -. - -

l Exhibit D Technical Specification Pages Marked Up with Proposed Changes j

i Monticello Nuclear Generating Plant -

Supplement 2 to License Amendment Request Dated August 15,1996 Exhibit B consists of the existing Technical Specification pages marked up with the proposed changes from submittals on August 15,1996, March 19,1998, and of Supplement 2. Pages I

- affected by this change are listed below; Page i, ii, iii, iv, v vi, vii, 22, 31, 61, 62 63a,69,72,82,89 99,102,125,126,164 188,190,198t, 200 223,227b,227c,227d,227e 229b, 229c, 229ff, 229i, 229u 232,233,234,237,239 240,241,242,243,244a,246b 247a,250 4

B-1 9810200125 981012 PDR ADOCK 05000263, P

PDR w

\\

%99 W WTr u

TABLE OF CONTENTS LiLLt 1.0 DEFINITIONS 1

2.0 SAFETt LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 6

2.1 and 2.3 Fuel Cladding Integrity 6

2.1 Bases 10 2.3 Bases 14 2.2 and 2.4 Reactor Coolant System 21 h

2.2 Bases 22 2.4 Bases 24 g

3.0 LIMITING CONDITIONS FOR OPERATION AND 4.0 SURVEILLANCE REQUIREMENTS 25a ty 4,0 Surveillance Requirements

[

3.1 and 4.1 Reactor Protection System 26 3.1 Bases 35 4.1 Bases L}

3.2 and 4.2 Protective Instrumentation 45 A.

Primary Containment Isolation Functions

45 B.

Emergency Core Cooling Subsystems Actuation 46 C.

Control Rod Block Actuation 46 D.

Other Instrumentation 46a E.

Reactor Building Ventilation Isolation and Standby Cas Treatment System Initiation 47 T.

Recirculation Pump Trip Initiation and Alternate Rod Injection Initiation 48 C.

Safeguards Bus Voltage Protection 48 11. Instrumentation for S/RV low Low Set Logic 48 1.

Instrumentation for Control Room Habitability Protection 48 t.

3.2 Bases 64 4.2 Bases 72 3.3 and 4.3 control Rod System 76 A.

Reactivity Limitations 76 B.

Control Rod Withdrawal 77 C.

Scram Insertion Times 81 D.

Control Rod Accumulators 82 E.

Reactivity Anomalies 83 F.

Scran Discharge Volume 83A C.

Required Action 83A 3.3 and 4.3 Bases 84 i

kendment NO. 30,37,0.65 MAY 3 01989

Sug%ys -

e.wsevsT E too c L

repAwrd onb L

^

ve 3.4 and 4.4 Standby Liquid. Control System 93 A.

System Operation 93

-B.

Boron Solution Requirements 95 C.

96 3.4 and.4.4 Bases 99 3.5 cnd 4.5 Core and Containment / Spray Cooling System 101 101

-l

'A.

ECCS Systeps_

103 S.

RHR Intertie Return Line Isolation Valves C.

Containment / Spray Cooling System 104 D.

RCIC 105 E.

Cold Shutdown and Refueling Requirements 106 F.

Recirculation System 107 3.5/4.5 Bases 110 3.6 and 4.6 Primary System Boundary 121 A.

Reactor Coolant Heatup and Cooldown 121 B.

Reactor Vessel Temperature and Pressure 122 1

C.

Coolant Chemistry 123 D.

Coolant Leakage 126-Safety / Relief Valves

'127

-E.

F.

Deleted C.

Jet Pumps 128 H.

Snubbers 129 j

3.6 and 4.6 Bases 145 l

3.7 and 4.7 Containment Systems 156 A.

Primary Containment 156 B.

Standby Cas Treatment System 166 C.

Secondary Containment 169 D.

Primary Containment Isolation Valves 170 E.

-Combustible Gas Control System 172 l

3.7 Bases 175 4.7 Bases 183 11 l'

Amendment No. 77, 79, 102

6 9pemcwr 9

/Ro chaw JrepriwTede.,-

i ody un 3.8 and 4.8 Radioactive Effluents 192 A.

Liquid Effluents 192 B.

Caseous Effluents 197 C.

Solid Radioactive Vaste 198e D.

Dose from All Uranium Fuel Cycle Sources 198f 3.8 and 4.8 Bases 198u 3.9 and 4.9 Auxiliary Electrical Systems 199 A.

Operational Requirements for Starcup 199 B.

Operational Requirements for Contit.ued operation 200 1.

Transmission Lines 200 2.

Reserve Transformers 201 3.

Standby Diesel Generators 201 4

Station Battery System 203 5.

24V Battery Systems 203 3.9 Bases 204 4.9 Bases 205 3.10 and 4.10 Refueling 206 A.

Refueling Interlocks 206 B.

Core Monitoring 207 C.

Fuel Storage Pool Water Level 207 D.

Movement of Fuel 207 E.

Extended Core and Control Rod Drive Maintenance 208 3.10 and 4.10 Bases 209 3.11 and 4.11 Reactor Fuel Assemblies 211 A.

Average Planar Linear Heat Generation Rate 211-B.

Linear Heat Generation Rate 212 C.

Minimum Critical Power Ratio 213 3.11 Bases 216 4.11 Bases 218 I

3.12 and 4.12 Sealed Source Contamination 219 A.

Contamination 219 B.

Records 221 3.12 and 4.12 Bases 222 iii Amendment No. 29, 102

a.

3.13 and 4.13 Fire Detection Protection Systems 223 A.

Fire Detection Instrumentation 223 B.

Fire Suppression Water System 224-C.

Hose Stations 226 D.

Yard Hydrant Hose Houses 227 E.

Sprinkler Systems 227a i

F.

Halon Systems 227b G.

Penetration Fire Barriers 227b H.

Alternate Shutdown Systam 227c 3.13 Bases 228 SS8h 4.13 Bases 3.14 and 4.14 Accident Monitoring Instrumentation 229a 3.15 and 4.15 Inservice Inspection and Testing 229f b o 00 A&k Mh.

229 3,15 and 4.15 Bases 6M 229 t

ipenmenta Monitoring Program Mg Sogg\\@e Cobcdow 3.16 and 4.16 Ra yd gp A.

% ;_in;; _..d Analysis 6 Gi 229h j

B.

Land use census 229j 229k C.

Interlaboratory comparison Pr to, 229t i

3.16 and 4.16 Bases 229u 3.17 and 4.17 Control Room Habitability OWTTok boom A. 4 Ventilation System 229u 229v f

B.9EmergencyFiltrationSystem l

229y 3.17 Bases 229z 4.17 Bases 230 5.0 DESIGN FEATURES 230 5.1 Site 230 5.2 Reactor 230 5.3 Reactor Vessel 230 5.4 Containment 231 5.5 Fuel Storage 231 5.6 Seismic Designs 232 6.0 ADMINISTRATIVE CONTROLS 232

'6.1 Organization 237 6.2 Review and Audit 243 6.3 Special Inspection and Audits 6,4 Action to be taken if a Safety Limit is Exceeded 243 244 6.5 Plant Operating Procedures 24 C dh 6.6 Plant Operating Records 248 6

ting Rey ig en poh

_.._a,

.~

w aa 3f3

.... ~ ggy 11; 57 L.

s.

x

.ty _.

OM E T

A e

[

LIST OF FTCURES Firure No.

Pace No.

3.4.1 Sodium Pentaborate Solution Volume-Concentration 97 Requirements 3.4.2 Sodium Pentaborate Solution Temperature Requirements 98

, g-

.Mq 0.5.1 Giur,1. Lvvy Cy.c tivu Sucv.111.uv.. iv..c/Fiv. Cui..

100 I /

%)W

'm _.n &

3.6.1 Core Beltline Operating Limits Curve Adjustment 133 vs. Fluence 3.6.2 Minimum Temperature vs. Pressure for Pressure 134 Tests 3.6.3 Minimum Temperature vs. Pressure for Mechanical 135 Heatup or Cooldown Without the Core Critical 3.6.4 Minimum Temperature vs. Pressure for Core 136 Operation 4.6.2 Chloride Stress Corrosion Test Results @ 500*F 137 3.7.1 Differential Pressure Decay Between the Drywell 191 and Wetvell 3.8.1 Monticello Nuclear Generating Plant Site Boundary 198g for Liquid Effluents 3.8.2 Monticello Nuclear Generating Plant Site Boundary 198h for Caseous Effluents V

AMENDMENT No. 9,35,41,11,11,79 APR 9 139)

' 8 - s-%

FbWpp\\t.MM $

LIST OF TABLES Table Nom Ufe 3.1.1 Reactor Protection System (Scram) Instrument Requirements 28 4.1.1 Scram Instrument Functional Tests - Minimum Functional 32 Test Frequencies for Safety Instrumentation and Control Circuits 4.1.2 Scram Instrument Calibration - Minimum Calibration 34 Frequencies for Reactor Protection Instrument Channels 3.2.1 Instrumentation that Initiates Primary Containment 49 Isolation Functions 3.2.2 Instrumentation that Initiates Emergency Core Cooling Systems 52 3.2.3 Instrumentation that Initiates Rod Block 3

3.2.4 Instrumentation that Initiates Reactor Building Ventilation 59 Isolation and Standby Gas Treatment System Initiation (1b 3.2.5 Instrumentation that Initiates a Recirculation Pump Trip 60 and Alternate Rod Injection I

3.2.6 Instrumentation for Safeguards Bus Degraded Voltage and 60a Loss of Voltage Protection W

3.2.7 Instrumentation for Safety / Relief Valve Low-Low Set Logic 60b s<L 60d GL-3.2.8 other Instrumentation 3

(/)

3.2.9 Instrumentation for Control Room Habitability Protection 60e 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod 61 Block and Isolation Instrumentation 132a h 65t" Snubber Visual Inspection Interval 1981 3.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 3.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198k 198m 4.8.1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.8.2 Radioactive Caseous Effluent Monitoring Instrumentation 198n Surveillance Requirements 4.8.3 Radioactive Liquid Waste Sampling and Analysis Program 198p 198s 4.8.4 Radioactive Caseous Waste Sampling and Analysis Program vi REV 132 7/15/92-

~\\ $~Q j.

J l

Swp@eAowT 4 LIST OF TABLES (continued) i

[

Table No.

Eage 3.13.1 Safety Related Fire Detection Instruments 427m

'AC/

3.14.1 Instrumentation for Accident Monitoring 229b i

4.14.1 Minimum Test and Calibration Frequency for Accident 229d Monitoring Instrumentation i

4.16.1 Radiation Environmental Monitoring Program (REMP) _

229-1 j

Sample Collection and Analysis 4.16.2 REMP - Maximum Values for the Lower Limits of Detectio (LQ 29q 4.16.3 REMP - Reporting Levels for Radioactivity Concentrations

-229s in Environmental Samples 6.1.1 Minimum Shift Crew Composition 236 bg eW\\td i

l l

vii REV -120 Uf2 9,'9"

8-6-%

juepkw..E 3 Bases 2.2:

The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is esserdial that the integrity of this system be protected by establishing a pressure limit to be observed for all o erating conditions and whenever there is irradiated fuel in the reactor vessel.

jgo The pressure safety limit of 1335 psig as measured in the vessel steam space is equivalent to 1375 psig at the lowest e vation of the reactor coolant system. The 1375 psig value was derived from the design pressures of the reactor pressure vessel, c olant piping, and recirculation pump casing. The respective design pressures are 1250 psig at 575cF,1148 psig at 562cF, and sig at 575 F. The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes:

ASME Boiler and Pressure Vessel Code Section lil-A for the pressure vessel, ASME Boiler and Pressure Vessel Code Section ill-C for the recirculation pump casing, and the USAS Piping Code Section B31.1 for the reactor coolant system pipir.g. The ASME Code permits pressure transients up to 10 percent over the vessel design pressure (110% x 1250 = 1375 psig) and the USAS Code permits pressure transients up to 20 percent over the piping design pressure (120% x 1148 = 1378 psig).

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig and temperature of 575oF; this is more than a factor of 1.5 below the yield strength of 42,300 psi at this temperature. At the pressure limit of 1375 psig, the general membrane stress increases to 29,400 psi, still safely below the yield strength.

The reactor coolant system piping provides a comparable margin of protection at the established pressure safety limit.

2.2 BASES 22 4/30/98 Amendment No. O,100a

L 4

Table 3.1.1 - Continued The high drywell pressure _sc_ ram functions in the Startup and Run modes' when necessary6niv_by closinst_ the e.

manuai containamsolation vagduring purging for containment inerting or de-inerting$ Verification N e bypass condition shall be noted in the control room log.

f.

0._ instrument channel for.the functions indicated in the table to allow completion of surveillance. testing, provided that:

4-1.

Redundant instrument channels in the same trip system are capable of initiating the automatic function and are demonstrated to be operable either immediately prior or_immediately subsequent to applying the bypass.

2.

While the bypass is applied, surveillance testing shall proceed on a continuous basis and the remaining instrument channels initiating the same function are tested prior to any other. Upon completion of surveillance testing, the bypass is removed.

t a

4 31 3.1/4.1 REVil 1/ y/ o r

{ _ g-__-_-_ -

z _,,..- m.

This page reflects changes proposed July 5, 1995 titled "Saveillance Test interval / Allowed Outage Tirne Extension Prograrn - Part 2.*

M x

Table 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, g

Rod Block and Isolation Instrumentation instrument Channel Test (3)

Calibration (3)

Sensor Check (3)

ECCS INSTRUMENTATION 1.

Reactor Low-Low Water Level Once/3 months (Note 5)

Every Operating Cycle -Transmitter l

Once/3 months -Trip Unit Onc^ m 2.

Drywell High Pressure Once/3 months Once/3 months Non I

3.

Reactor Low Pressure (Pump Start)

Once!3 months Once/3 months None 4.

Reactor Low Pressure (Valve Once/3 months Once/3 months None Permissive) 5.

Undervoltage Emergency Bus Refueling Outage Refueling Outage None 6.

Low Pressure Core Cooling Pumps Once/3 months Once/3 months None l

Discharge Pressure Interlock 7.

Loss of Auxiliary Power Refueling Outage Refueling Outage None 8.

Condensate Storage Tank Level Refueting Outage Refueling Outage Non 7-- -,

9.

Reactor High Water Level Once/3 months (Note 5)

Every Operating Cycle - Transmitter l1 bo* C l Every 3 months - Trip Unit On Q"

,F ROD BLOCKS 1.

APRM Downscale Once/3 months (Note 5)

Once/3 months None 2.

APRM Flow Variable Once/3 months (Note 5)

Once/3 months None 3.

IRM Upscale Notes (2,5)

Note 2 Note 2 4.

IRM Downscale Notes (2,5)

Note 2 Note 2 5.

RBM Upscale Once/3 months (Note 5)

Once/3 months None 6.

RBM Downscale Once/3 months (Note 5)

Once/3 months None 7.

SRM Upscale Notes (2,5)

Note 2 Note 2 8.

SRM Detector Not-Full-in Position Notes (2,9)

Note 2 None 9.

Scram Discharge Volume-High Level Once/3 months Refueling Outage None MAIN STEAM LINE (GROUP i) ISOLATION 1.

Steam Tunnel High Temperature Refueling Outage Refueling Outage None 7 hewe,)

2.

Steam Line High Flow Once/3 months Once/3 Months Oncq""M"

,s v

I 3.2/4.2 61 Amendment No. 2,40,37,39,63,66,81 i

l

3-M-%)

i M.MS This page reflects changes proposed July 5, 1995 titled " Surveillance Test Interval / Allowed ng 9 Outage Time Extension Program - Part 2."j

(,

e Table 4.2.1 Continued Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation instrumentation

[

Instrument Channel Test (3)

Calibration (3)

Sensor Check (3) 3.

Steam Line Low Pressure Once/3 months Once/3 months None Onc['.riewes j 4.

Reactor Low Low Water Level Once/3 months (Note 5)

Every Operating Cycle-Transmitter l2 Once/3 Months-Trip Unit CONTAINMENT ISOLATION (GROUPS 2 & 3}

1.

Reactor Low Water Level (Note 10) 2.

Drywell High Pressure (Note 10)

HPCI (GROUP 4) ISOLATION 1.

Steam Line High Flow Once/3 months Once/3 months None 2.

Steam Line High Temperature Once/3 months Once/3 months None RCIC (GROUP 5) ISOLATION 1.

Steam Line High Flow Once/3 months Once/3 months None 2.

Steam Line High Temperature Once/3 months Once/3 months None REACTOR BUILDING VENTILATION & STANDBY GAS TREATMENT 1.

Reactor Low Low Water Level Once/3 months (Note 5)

Every Operating Cycle - Transmitter Onc'.

Once/3 months - Trip Unit 2.

Drywell High Pressure (Note 10) 3.

Radiation Monitors (Plenum)

Once/3 months Once/3 months Once/ day 4.

Radiation Monitors (Refueling Floor)

Once/3 months Once/3 months Note 4 RECIRCULATION PUMP TRIP AND ALTERNATE ROD INJECTION 1.

Reactor High Pressure Once/3 months (Note 5)

Once/ Operating Cycle-Transmitter Once/Dav g

Once/3 Months-Trip Unit 2,

arg,

2.

Reactor Low Low Water Level Once/3 months (Note 5)

Once/ Operating Cycle-Transmitter On ^/ch fe Once/3 Months-Trip Unit SHUTDOWN COOLING SUPPLY ISOLATION 1.

Reactor Pressure interlock Once/3 months Once/3 Months None l

3.2/4.2 62 Amendment No. 74,84,83,91

3 -M -% 8 -

Table 4.2.1 - Continued Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation. Instrumentation NOTES:

(1)

(Deleted) ggQg (2)

Calibrate prior to normal shutdown and start-up and thereafter check once per -hi't.nd test once per week until no longer required. Calibration of this instrument prior to normal shutdown means adjustment of channel trips so that they correspond, within acceptable range and accuracy, to a simulated signal injected into. the instrument (not primary sensor).

In addition, IRM gain adjustment will be performed, as necessary, in the.

APRM/IRM overlap region.

(3)

Functional tests, calibrations and sensor checks are not required when the systems are not required to be operable or are tripped. If tests ari missed, they shall be performed prior to returning the systems to an operable status.

g7

DWtS, (4)

Whenever fuel handling is in process, a sensor check shall be performed once per i f t-g (5)

A functional test of this instrument means the injection of a simulated signal into the instrument (not primary sensor) to verify the proper instrument channel response alarm and/or initiating action.

(6)

(Deleted)

(7)

(Deleted)

(8)

Once/ shutdown if not tested during previous 3 month period.

(9)

Testing of the SRM Not-Full-In rod block is not required if the SRM detectors are secured in the full-in position.

(10) Uses contacts from scram system. Tested and calibrated in accordance with Tables 4.1.1 and 4.1.2.

3.2/4.2 63a REV-133 S/1?/92-

7\\

Q g b pp h g r ft Bases 3.2 (Continuedh increases core voiding, a negative reactivity feedback. High pressure sensors initiate the pump trip in the event of an isolation.

transient. Low level sensors initiate the trip on loss of feedwater (and the resulting MSIV closure). The recirculation pump trip is only required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam which continues to be generated after reactor isolation in this unlikely postulated event, requiring the trip to be operable only when in the

. RUN mode is therefore conservative.

The ATWS high reactor pressure and low-low water levellogic also initiates the Attemate Rod injection System. Two solenoid valves are installed in the scram air header upstream of the hydraulic control units. Each of the two trip systems energizes a valve to vent the header and causes rod insertion. This greatly reduces the long term consequences of an ATWS event.

Voltage sensing relays are provided on the safeguards bus to transfer the bus to an attemate source when a loss of voltage condition or a degraded voltage condition is sensed. On loss of voltage this transfer occurs immediately. The transfer on degraded voltage has a time delay to prevent transfer during the starting of large loads. The degraded voltage setpoint corresponds to the minimum acceptable safeguards bus voltage for rt f.;, rd --M; ' r" it; e '^rr Of ;;de.J ;;ddd An allowance for relay tolerance is included.

Safety / relief valve low-low set logic is provided to prevent any safe ef valve from opening when there is an elevated water leg in the respective discharge line. A high water leg is formed immediat following valve closure due to the vacuum formed when steam condenses in the line. If the valve reopens before the dischar line vacuum breakers act to retum water level to normal, water clearing thrust loads on the discharge line may exceed their de sign limit. The logic reduces the opening setpoint and increases the blowdown range of three non-APRS valves following scram. A 15-second interval between subsequent valve actuations is provided assuming one valve fails to x

a sTe L ied rbr e r A de g t

stee LocA M0 T. # N C-9 DOV e.%c.yCflod 1% C.(, g,

A A

1 3.2 BASES 69

'4/30/98 Amendment No. 30,34,45,100a i

=

3 9 s 6wppbw:wT 3.

Bases 4.2:

The instrumentation in this section will be functionally tested and calibrated at regularly scheduled intervals. Although this instrumentation is not generally considered to be as important to plant safety as the Reactor Protection System, the same design reliability goals are applied. As discussed in Section 4.1 Bases, monthly or quarterly testing is generally specified unless the testing must be conducted during refueling outages. Quarterly calibration is specified unless the calibration must be conducted during refueling outages. Where applicable, sensor checks are specified on a once/ shift.or ons/ day basis.

12. bouts d nce 4.2 BASES 72 4/30/98 Amendment No. 63,84,100a

nw g~-

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIRiCE REQUIREMENTS Any four rod group may contain a control

/' ~

~~

~

C T'

\\2L I E} ('bY rod which is valved out of service provided the above requirements and Specification 3.3.A are met.

D.

Control R.

Accumulators D.

Control Rod Accumulators Once e-shift heck the status in the con rol room of the Control rod accumulators shall be operable in required Operable accumulator the Startup, Run, or Refuel modes except as pressure and level alares, provided below.

1.

In the Startup or Run Mode, a rod accumulator may be inoperable provided that no other control rod in the nine-rod square array around this rod has a:

(a)

Inoperable accumulator, or (b)

Directional control valve electrically disarmed while in a non-fully inserted position.

If a control rod with an inoperable accumulator is inserted " full-in" and its directional control valves are electrically disarmed, it shall not be considered to have an inoperable accumulator.

82 3.3/4.3 REV-113 4/IS/92

1pplomcwT2 Bases 3.3/4.3 (Continuedh consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second -

assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous pattems of scattered control rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism.

C.

Scram insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the Safety Limit (T.S.2.1.A). This requires the negative reactivity insertion in any local region of the core and in the overall core to be equivalent to at least the scram reactivity curve used in the transient analysis. The required average scram times for three control rods in all two by two arrays and th0 required average scram times for all control rods are based on inserting this amount of negative reactivity at the specified rate locally and in the overall core. Under these conditions, the-therme! 'imite are never reached duMg the trenciente requirMg COMre! red Ocram e !t-itMg eporatione! trencier! !c the! reeu!!'ag n

from a tutine !Op v0!v0 O!Ocurc "" failure of th0 tutin0 byp000 cyct0m fa0!yci Of th? tran0!Ont ch0v/0 that th0 negativo re=c'Iuity ra'ee resulting frem +he serem -ith the everege reepenee of e!! +5e driv 0: et given in the ebeve Specificat!On, provido the rna"irad pra'ac+ien, and MCPR remains above the Safety Limit (T.S.2.1.A).

CM. ve

& c.eg

%4c) bit \\ s we er ucedel Adg sE C.oWTTok to S C.Ta.M

&]

~T % dewy r

y A

3.3/4.3 BASES 89 4/30/98 Amendment No. 29,100a

~~

\\B-s-%

h 4%e%Yh a k % \\eo u ie A % ea4-a & &. 2 ro F

Bases 3.4/4.4:

l A.

The design objective of the standby liquid control system is to provide it e capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rod can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a co entration of boron in the reactor core in less than 125 minutes sufficient to bring the reactor from full power to a 3% delta k subcri ' al condition considering the hot to cold reactivity swing, xenon poisoning and an additional 25% boron concentration mar in f~ ecccib!c impc"cct rnixing Of '50 chemicc! cc!ution in the rocctw wcter Ond diluticr 'rc~' '50 ">cter "'50 coc!dev~' circu?!

The time requirement (125 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.

The ATWS Rule (10 CFR 50.62) requires the addition of a new design requirement to the generic SLC System design basis.

Changes to flow rate, solution concentration or boron enrichment to meet the ATWS Rule do not invalidate the original system design basis. Paragraph (c)(4) of 10 CFR 50.62 states that:

"Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution" (natural boron enrichment).

The described minimum system parameters (equivalent to 24 gpm,10.7% concentration and 55 atom percent Boron-10 enrichment) will ensure an equiva ent injection capability that meets the ATWS rule requirement.

Boron enrichment concentration, solution temperature, and volume (including check of tank heater and pipe heat tracing system) are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. A reliability analysis indicates that the plant can be operated safely in this manner for ten days. For additional margin, the allowable out of service time has been reduced to seven days.

The only practical time to test the standby liquid control system is during a refueling outage and by initiation from local stations.

Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the replacement charges for the tested system are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control room.

The relief valves in the standby liquid control system protect the system piping and positive displacement pumps which are nominally designed for 1500 psi from overpressure. The pressure relief valves discharge back to the standby liquid control solution tank.

3.4/4.4 BASES 99 4/30/98 Amendment No. 56,57,-7-7,100a

J 4.0 SURVEILLANCE REQUIREMENTS 3.0 LIMITING CONDITION FOR OPERATION 4.

Perform the following tests:

3.

One of the following conditions of inoperability may exist for the period I

Item Frecuency specified:

One Core Spray subsystem may be inoperable Motor Operated Pursuant to Valve Operability Specification a.

for 7 days, or b.

One RHR pump may be inoperable for 30 days, r

ADS Valve Each Operating

)

Operability Cycle One low pressure pump or valve (Core Spray

)

c.

or RHR) may be inoperable with an ADS valve Note: Safety / relief valve operability is

[

inoperable for 7 days, or verified by cycling the valve and observing

)

d.

One cf the two LPCI injection paths may be a compensating change in turbine bypass inoperable for 7 days, or valve position.

OC C.O m ro Two RHR pumps may be inoperable for 7 days, ADS Inhibit Each Operating e.

or Swit h Operability Cycle f.

Both of the LPCI injection paths may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or Perform a simulated Each Operating

(

automatic actuation test Cycle g.

HPCI may be inoperable for 14 days, (including HPCI transfer to provided RCIC is operable, or P ress poo and

(

[ h.

One ADS valve may be inoperable for 14 days, or subsequent low reactor 1.

Two or more ADS valves may be inoperable water level) for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5.

Perform the following test on the Core 4.

If the requirements or conditions of 3.5. A.1, Spray Ap Instrumentation:

2 or 3 cannot be met, an orderly shutdown of

\\

the reactor shall be initiated and the reactor Check Once/ day j

shall be placed in a condition in which the affected equipment is not required to be Test once/ month operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Calibrate Once/3 months 102 3.5/4.5 REV 129 ^/9/91

(3 -a C

_t 1

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 2.

(a) The reactor coolant water shall not 2.

During startup and at steaming rates exceed the following limits with below 100,000 pounds per hour, a sample steaming rates less than 100,000 of reactor coolant shall be taken every pounds per hour except as specified four hours and analyzed for conductivity and chloride content.

in 3.6.C.2.b.

Conductivity 5 umho/cm Chloride ion 0.1 ppm (b) For reactor startups the maximum value for conductivity shall not exceed 10 umho/cm and the maximum value for chloride ion concentration shall not exceed 0.1 ppm for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after placing the reactor in the power operating condition, the 3.

(a) With steaming rates greater than b.

Except as specified in 3.6.C.2.b above, reactor coolant kater shall not exceed the or equal to 100,000 lbs'. per hour, a reactor coolant sample shall be taken following limits with steamfng rates greater at least every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and when the than or equal to 100,000 lbs. per hour.

continuous conductivity monitors in-dicate abnormal conductivity (other Conductivity 5 umho/cm than short-term spikes) and analyzed Chloride ion 0.5 ppm for conductivity and chloride ion content.

4.

If Specifications 3.6.C.1 through 3.6.C.3 are not met, an orderly shutdown shall be initiated and the reactor shall be in the (b) When the continuous conductivity 6

cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

monitor is inoperable, during power h, g operation, a reactor coolant sam ge

[5r /

should be taken and analyzed for coHTuTtIVITy

^

r an chloride ion cotttent.

7 w.fs 125 3.6/4.6

-ncz 52 1/9 fat

(3-6-%

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 -SURVEILLANCE REQUIREMENTS m

D.

Coolant Leakage D.

Coolant Leakage '

1.

Any time irradiated fuel is in the 1.

Any time irradiated fuel is.in the reactor reactor vessel and coolant temperature is vessel and coolant temperature is above above 212'F, reactor coolant system 212'F, the following surveillance program leakage, based on mump monitoring, shall shall be carried out ghte$g.,

be limited to

~

a.

Unidentified and Identified Leakage a.

5 gpm Unidentified Leakage rates shalA_be recorded once per b.

2 gpm increase in Unidentified

2 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months using prim y--v Leakage within any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period at floor and equipment drain c.

20 gpm Identified Leakage sump monitoring equipment, d.

no pressure boundary leakage 2.

The reactor coolant system leakage 2.

With reactor coolant system leakage detection systems shall be demonstrated greater than 3.6.D.1.a or 3.6.D.1.c OPERABLE by above, reduce the' leakage rate to within a.

Primary containment atmosphere acceptable limits within four hours or initiate an orderly shutdown of the Particulate monitoring systems,, t reactor and reduce reactor water Performance of a sensor check _3,1rrr],]

g temperature to less than 212*F within 24 once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a channel

hours, functional test at least monthly and a channel calibration'at least once per 3.

With an increase in Unidentified Leakage cycle.

in excess of the rate specified in b.

Primary containment sump leakage 3.6.D.1.b, identify the source of a ste -pegformance mea nt a

increased leakage within four hours or nce per -ehtfe initiate an orderly shutdown of the sensog,,chec

_t _ @ hours and a channet 12-reactor and reduce reactor water e-c-temperature to less than 212'F within 24 hQgCE cal ation est at least once per cycle.

hours.

t 4.

If any Pressure Boundary Leakage is l

detected when the corrective actions outlined in 3.6.D.2 and 3.6.D.3 above are taken, initiate an orderly shutdown of l

the reactor and reduce reactor water temperature to less than 212*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

126 i

3.6/4.6 REV 1??- ',' C / Q A i

i s

i a

'LM 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS

4. Pressure Suppression Chamb r-prywell_V cuum 4.

Pressure Suppression Chamber-Drywell Vacuum 4

Breakers

~

Breakers 7

Operability and full closure of the When primary containment is requ red, all a.

eight drywell-suppression chamber vacuum drywell-suppression chamber vacuum a.

breakers shall be operable and positioned breakers shall be verified by performance in the closed position as indicated by of the following:

the position indication system, except during testing and except as specified in (1)

Monthly each operable drywell-3.7.A.4.b through 3.7.A.4.d below.

suppression chamber vacuum breaker i

shall be exercised through an opening-closing cycle.

b.

Any drywell-suppression chamber vacuum breaker may be nonfully closed as indicated by the position indication and (2)

Once each operating cycle, dry-well to suppression chamber i

alarm system provided that drywell to leakage shall be demonstrated to suppression chamber differential pressure be less than that equivalent to a decay does not exceed that shown on one-inch diameter (zifice and each Figure 3.7.1 vacuum breaker shall be visually inspected.

(Containment access Up to two drywell-suppression chamber required) c.

vacuum breakers may be inoperable provided that: (1) the vacuum breakers are determined to be fully closed and at (3)

Once each operating cycle, vacuum least one position alarm circuit is breaker position indication and operabic or (2) the vacuum breaker is alarm systems shall be calibrated and functionally tested.

secured in the closed position or (Containment access required) replaced by a blank flange.

(4)

Once each operating cycle, the d.

Drywell-suppression chamber vacuum vacuum breakers shall be tested to breakers may be cycled, one at a time, determine that the force required

_~

' lIT during containment inerting and to open each valve from fully deinerting operations to assist in closed to fully open does not purging air or nitrogen from the exceed that equivalent to 0.5 psi suppression chamber vent header.

acting on the suppression chamber face of the valve disc.

(Containment access required) 164 REV-13^

S/12,'^1-3.7/4.7 i

\\

-s- %

8 S-@%

Bases 4.7 (Continuedt B.

Standby Gas Treatment System, and C. Secondary Containment initiating reactor building isolation and operation of the standby gas treatment system to maintain the design negative pressure within the secondary containment provides an adequate test of the reactor building isolation valves and the standby gas treatment system. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system operational capability. Secondary Containment Capability Test data obtained under non-calm conditions is to be extrapolated to calm wind conditions using information provided in " Summary Technical Report to the United States Atomic Energy Commission, Directorate of Licensing, on Secondary Containment Leak Rate Test", submitted by letter dated July 23,1973, and as described in NSP letter to the NRC dated August 18,1995, with subject," Revision 2 to License Amendment Request Dated June 8,1994, Standby Gas Treatment and Secondary Containment Technical Specifications."

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Standby gas treatment system inplace testing procedures will be established utilizing applicable sections of ANSI N510-1989 standard as a procedural guideline only. If painting, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis should be performed as required for operational use. Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52 Revision 2 (March 1978 The charcoal adsorber efficiency test procedures will allow for the removal of a representative sample. The 30*C,95% relativ humidity test per ASTM D 3803-89 is the test method to establish the methyliodine removal efficiency of adsorbent. The sa le will be at least two inches in diameter and a length equal to the thickness of the bed. If the iodine removal efficiency test r suits are unacceptable, all adsorbent in the system will be replaced.

High efficiency particulate filters are installed before and ter the charcoal filters to prevent clogging of the carbon adsorbers and to minimize potential release of particulates to the e ironment. An efficiency of 99% is adequate to retain particulates that may be released to the reactor building following an ac dent. This will be demonstrated by inplace testing with DOP as the testing medium. Any HEPA filters found defective will replaced with filters qualified pursuant to regulatory guide position C.3.d of Regulatory Guide 1.52 Revision 2 (March 197 ). Once per operating cycle demonstration of HEPA filter pressure drop, operability of inline heaters at rated power, automatic nitiation of each standby gas treatment system circuit, cad leakage tests after maintenance or testing which could affect leak ge, is necessary to assure system performance capability.

C%CCf TC$T dcM be DW MM D M M -- B R WV 8

4/30/98 4.7 BASES Amendment No. 94,100a

S weg k m. c J 1

Bases 4.7 (Continuedh The containment is penetrated by a large number of small diameter instrument lines. A program for the periodic testing (see Specification 4.7.D) and examination of the valves 'n se lines has been developed and a report covering this program was submitted to the AEC on July 27,1993.

{ ]

The main steam line isolation valves are functs s ed on a more frequent interval to establish a high degree of reliability.

E.

Combustible Gas Control System The Combustible Gas Control System (CGCS) is functionally tested once every six months to ensure that the recombiner trains will be available if required. In addition, calibration and maintenance of essential components is specified once each operating cycle.

4.7 BASES 190 4/30/98 Amendment No. 36,100a

Mwa TABLE 4.8.4

- RADI0 ACTIVE CASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (continued)

(Page 2 of 2)

Notes:

The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95%

a.

probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

Note (a) of Table 4.8.3 is applicable.

b.

Crau samples taken at the discharge of the plant stack and reactor building vent are generally below minimum detectable levels for most nuclides with existing analytical equipment.

For this reason, isotopic analysis data, corrected for holdup time, for samples taken at the steam jet air ejector may be used to calculate noble gas ratios c.

Whenever the steady state radiciodine concentration is greater than 10 percent of the limit of Specification 3.6.C.1, daily sampling of reactor coolant for radioactive iodines of I-131 through I-135 is required.

Whenever a change of 25% or more in calculated Dose Equivalent I-131 is detected under these conditions, the iodine and particulate collection devices for all release points shall be removed and analyzed daily until it is shown that a pattern exists which can be used to predict the release rate.

Sampling may then revert When samples collected for one da are analyzed, the corresponding LLD's may be increased by a to weekl.10.

Samples shall be analyzed with n 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after removal factor o d.

To be representative of the average quantities and concentrations of radioactive materials in particulate form in gaseous effluents, samples shoul be collected in proportion to the rate of flow of the effluent gQg ggg g streams.

e.

The p ipal gamma emitters for which the LLD specification will apply are exclusively the following radi uclides:

Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, co-60, 2n-65,hese nuclides a,reKr-87, Kr-88, Xe-133, Xe-133m,141,d Mo-99, Cs-134 Cs-137, Ce-and Ce-144 for particulate emissions. This list does not mean that only t to be detecte and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

f.

Nuclides which are below the LLD for the analyses shall be reported as "less than" the LLD of the nuclide and should not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations. When unusual circumstances result in LLD's higher than reported, the reasons shall be documented in the semiannual effluent report, g.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period sampled.

if the limits of 3.8.B.1 are satisfied for other 113 analysisshallnotbegequiredpriortopurginfetedwithin24hoursaftersampling.

nuclides.

Iloweve r, the 11 analysis shall be comp 1.

In lieu of grab samples, continuous monitoring with bi-weekly analysis using silica-el samplers may be provided.

198t 3.8/4.8 REV 1'O 0/7/0'

bMSM 3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 2.

Both diesel generators are operable and capable of feeding their designated 4160 volt buses.

ft' 3.

(a) 4160V Buses #15 and #16 are energized.

f (b) 480V Load Centers #103 and #104'are energized.

4.

All station 24/48, 125, and 250 volt batteries are charged and in service, and associated battery chargers are operable.

B.

When the mode switch is in Run, the availability of electric power shall be as specified in 3.9. A.

except as specified in 3.9.B,1, 3.0.0.2, 0.0.0.0-2nd _. 9. ". '- or the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1.

Transmission Lines From and after the date that incoming power is available from only one line, reactor operation is permissible only during the succeeding seven days unless an additional line is sooner placed in 200 3.9/4.9 REV 101 10/10/07 i

'3 -$12 h

\\

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILlRICE REQUIREMENTS 3.13 FIRE DETECTION AND PROTECTION SYSTEMS 4.13 FIRE DETECTION AND PROTECTION SYSTEMS Applicability:

Applicability:

Applies to instrumentation and plant systems used for Y pplies to the periodic testing of instru-fire detection and protection of the nuclear safety-mentation and plant systems used for fire related structures, systems, and components of the pla detection and protection of the nuclear safety related structures, systems, and Obiective:

components.

To insure that the structures, systems, and components Obiective:

of the plant important to nuclear safety are protected from fire damage.

To verify the operability of instrumentation and plant systems used for fire detection and Specification:

protection of nuclear safety related structures, systems, and components.

A.

Fire Detection Instrumentation Specification:

1.

Except as specified below, the minimum fire detection instrumentation for each fire A.

Fire Detection Instrumentation detection zone shown in Table 3.13.1 shall be operable whenever equipment in that 1.

Fire detection instrumentation in each of fire detection zone is required to be the zones in Table 3.13.1 shall be demon-operable.

strated operable every six months by performance of functional tests.

2.

If specification 3.13.A.1 cannot be met, within one' hour establish a fire watch patrol to inspect 2.

Ala e4 = 'n associated with the fire the zone (s) with inoperable instruments C_:::-

once per hour Restore the minimum numb M detectorp nstruments in each of the zones m

in Table '3.13.1 shall be demonstrated instruments o operable status within 14 days or operable every six months, submit a sp cial report to the Commission within 30 days ou lining the cause of the inoperability ekeg gM'g' and the p ans and schedule for restoring the instrume es to operable status.

u

+ 2. 5 2, 3.13/4.13 223 7lL &6 mc: %

i

3-W%

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS F.

Halon Systems F.

Halon Systems 1.

The cable spreading room Halon system shall 1.

The cable spreading room Halon system be operable with the storage tanks having shall be demonstrated operable as follows:

r e wei ht and 90% of at M 951 of JuQ c 6

u C hh kl% Qn=1, pe"e_ erMC' 3e\\[

a.

Each valve (manual, power operated, or lun en-lig'eTsisure?

a automatic) in the flow path that is not 2.

If specification 3.13.F.1 cannot be met, electrically supervised, locked, within one hour establish a continuous fire sealed or otherwise secured in position, watch with backup fire suppression equipment shall be verified to be in its correct in the cable spreading room.

Restore the position every month, system to operable status within 14 days or submit a special report to the Commission b.

Verify Halon storage tank weight and within 30 days outlining the cause of the pressure every six months.

inoperability ar.d the plans and schedule a.

Perform a system functional test every for restoring the system to operable status.

c.

18 months which includes verifying the G.

Fenetration Fire Barriers system, including associated ventilation dampers, actuates manually and automati-1.

All penetration fire barriers in fire area cally, upon receipt of a test signal boundaries shall be operable whenever safe shutdown equipment in that fire a d.

Perform an air flow test every 3 years G(.

c) through headers and nozzles to assure required to be operable.

no blockage.

2.

If Specification 3.13.G.1 ca ot be B t, Visually examine headers and nozzles a continuous fire watch sha I be e.

established on at least on side of the every 18 months. An air flow test shall affected penetration (s thin one hour or be performed upon evidence of obstruc-verify the operabilit ire detectors on tion of any Halon system nozzle.

at least one side of tie non-functional fire barrier and establish an hourly fire watch G.

fenetration Fire Barriers patrol. Restore the inoperable penetration fire barriers to Operable stat s within 14 1.

A visual inspection of penetration fire days or submit a special repor to the barriers in fire area boundaries protecting

,a Commission within 30 days out1 ning the safe shutdown equipment shall be conducted cause of the inoperability an the plans and every 18 months.

r schedule for restoring the bar iers to Operable status.

2.

Following repair or maintenance of a pene-tration fire barrier a visual inspection l 7,f[j]s}

of the shall be conducted.

227b 3.13/4.13 Amendment No. 7,M,61 x

.m s>

L a 4 e /,5

.p-

f3-15ic14 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 11. Alternate S1utdown System 11. Alternate Shutdown System 1.

The system controls on the ASDS pane 5 $ Tens /ca*Pom*5' 1.

Switches on the alternate shutdown system

/

panel shall be functionally tested once hall be operable whenever that :ble.

al-ere required to be opera per operating cycle.

15 her 9a aaeel reem 2.

The alternate shutdown system panel master 2.

If system controls required to be transfer switch shall be verified to alarm operable by Specification 3.13.11.1 are in the control room when unlocked once per made or found inoperable, restore the operating cycle.

inoperable system control to operable within 7 days, or perform one of the

/

following; p/

f a.

Provide equivalent shutdown capability

/

j and within 60 days restore the inoperable j

system controls to operable; or

(

b.

Establish a continuous fire watch in the i

cable spreading room and the back-panel

)

area of the control room and within 60 days restore the inoperable system controls to

/

operable; or c.

Verify the operability of the fire detectors in the cable spreading room and the back-panel area of the control room and establish a hourly fire watch patrol and within 60 days restore the inoperable system controls to operable; or d.

Place the reactor in a condition where the systems for which the system controls at the ASDS are inoperable are not required to be operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l 3.

The alternate shutdown system panel master transfer switch shall be locked in the normal position except when in use, being tested or being maintained.

3.13/4.13 227c REV -111 b29 uv

/

$6-\\5-4B J

TABLE 3.13.1

[O*

SAFETY REIATED FIRE DETECTION INSTRUMENTS ge.u Y

Minimum Instruments Operable y

Heat Flame Smoke Fire Zone Location 3

1A "B" RHR Room 3

1B "A" RHR Room 3

1C RCIC Room 2

1E HPCI Room 11 IF Reactor Building-Torus Compartment 1

2A Reactor Bldg. 935' elev - TIP Drive Area 10.

2B Reactor Bldg. 935' elev - CRD HCU Area East 11 2C Reactor Bldg. 935' elev - CRD HCU Area West 1

+

AH Reactor Bldg. 935' - LPCI Injection Valve Area 2

3B Reactor Bldg. 962' elev - SBLC Area 5

3C Reactor Bldg. 962' elev - South 4

3D Reactor Bldg. 962' elev - RBCCU Pump Area 4

4A Reactor Bldg. 985' elev - South 5

4B Reactor Bldg. 985' elev - RBCCU Hx Area 2

4D SBGT System Room 7

SA Reactor Bldg. 1001' elev - South 3

SB Reactor Bldg. 1001' elev - North 1

SC Reactor Bldg. - Fuel Pool Cooling Pump Area 5

6 Reactor Building 1027' elev 1

7A Battary Room 1

7B Battery Room 1

7C Battery Room 7

8 Cable Spreading Room i

227d 3.13/4.13 REV 111 3/29/89

-W-%

v s_

-l

~;mcwT ge.N'CD TABLE 3.13.1 SAFETY RELATED FIRE DETECTION INSTRUMENTS p\\e,mSM Minimum Instruments Operable g

Heat Flame Smoke Fire Zone Location 12A Turbine Bldg. - 911' - 4.16 KV Switchgear 3

13C Turbine Bldg. - 911'

- MCC 133 Area 1

14A-Turbine Bldg. - 931' - 4.16.KV Switchgear 2

15A//sde

12 DG Room & Day Tank Room 3

ISB/jqp

11 DG Room & Day Tank Room 3

16 Turbine Bldg. 931' elev - Cable Corridor 3

17 Turbine Bldg. 941' elev - Cable Corridor 3

19A Turbine Bldg. 931' elev - Water Treatment Area 5

19B Turbine Bldg. 931' elev - MCC 142-143 Area 1

19C Turbine Bldg. 931' elev - FW Pipe Chase 1

20 Heating Boiler Room 1

23A Intake Structure Pump Room 3

31A ist Floor - Reactor Building Addition - Division I 3-31B 1st Floor - Reactor Building Addition - Division II 15 32A 2nd Floor - Reactor Building Addition - Division 1 6

32B 2nd Floor - Reactor Building Addition - Division II 4

/

33 3rd Floor - Reactor Building Addition 5

(

3.13/4.13 227e REV 111 ?/29/89

h Table 3.14.1 Instrumentation for Accident Monitoring Function Total No. of Minimum No. of Required Instrument Channels Operable Channels Conditions

Reactor Vessel Fuel Zone Water Level 2

1

. A, B Safety / Relief Valve Position 2

1 A, C (One Channel Pressure Switch and One Channel Thermocouple Position Indication per Valve)

Drywell Wide Range Pressure 2

1 A, B Suppression Pool Wide Range Level 2

1 AB Suppression Pool Temperature 2

1 A, D

{

Drywell High Range Radiation 2

1 A, D Drywell and Suppression Pool 2

1 A, B Hydrogen and Oxygen Monitor Offgas Stack Wide Range Radiation 2

1 A,

D Reactor Bldg Vent Wide Range Radiation 2

1 A, D

Required Conditions A. When the number of channels made or found to be inoperable is such that the number of operable channels is less than the total number of channels, either restore the inoperable channels to operable status within seven days, or prepare and submit a special report to the Commission pursuant to Technical Specification 5.'.S.2 within the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status.

L. J. h 3.14/4.14 229b REV-113 ' /10/0 ^-

3 -%-%

s

'?

Table 3.14.1 (contin'ued)

Instrumentation for Accident Monitoring

Required Conditions (continued)

B.

When the number of channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, the minimum number of channels shall be restored to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C.

When the number of channels made or found to be inoperable is such that the number of operable number of operable channels shown, the torus temperature shall channels is less than the mini be monitored (-~ 'e_ct sace pemhi-f to observe any unexplained temperature increase which might r

be indicative of an p3en SR ;

ie minimum number of channels shall be restored to operable status within 30 days or b&sLn at least flot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Cold Shutdown within

^.

the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D.

When the number of channels made or found to be inoperable is such that the number of operable channels is less then the minimum number of operable channels shown, initiate the preplanned alternate method of onitoring the appropriate parameters in addition to submitting the report required in (A) abov on ce. pu

12. hoes +26%

r 229c 3.14/4.14 REV -113

'. f 10,'0 %

-W-%

s_-

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS n-umam p

2/ Welds i austenitic painless steel pipi four 1.ches or larg t in diameter cont ning reactor coolant a a

tem erature abov 200 degre s F during

(

\\ eT e -

7p er operatio includin reactor ves 1

)

tachments - d appurte nces, shall e included i n augment inspection rogram meeting t requireme s of Gener Letter 88-01.

B.

Inservice Testing 1.

Inservice Testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with the requirements for ASME Code Class 1, 2 and 3 pumps and valves, respectively, contained in Section XI of the ASME boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) except where relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55(a)(g)(6)(1), or where alternate testing is justified in accordance with Generic Letter 89-04.

2.

Nothing in the ASME Boiler and Pressure

/

Vessel code shall be construed to supersede the requirements of any Technical Specification.

3.15/4.15 229ff REV -127 2/15/91

8-\\sA

-s s,

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS 3.

Deviations are permitted from the required sampling schedule if samples are unobtainable due to hazardous conditions, seasonable unavaila-bility, or to malfunction of automatic sampling equipment.

If the latter occurs, every effort shall be made to complete corrective action prior to the end of the next sampling period.

4.

With the level of radioactivity in an environ-mental sampling medium exceeding the reporting-levels of Table 4.16.3 when averaged over any calendar quarter, submit a special report to the Commission within 30 days from the end of the

/

affected calendar quarter pursuant to Specification 5

'.c.?.

When more than one of

(

the radion s in Table 4.16.3 are detected in ampling medium, this report shall be submitted if:

(j[,,h L

concentration (1) + concentration (2) + *** > 1.0 limit level (1) limit level (2)

When radionuclides other than those in Table 4.16.3 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.8.A.2, 3.8.B.2, or 3.8.B.3.

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiation Environmental Monitoring Report.

2291 3.36/4.16 REV-90 7/1/0G

3-W_gg 1

3.0 LIMITING CONDITIONS FOR OPERATION

- 4.0 SURVEILIANCE REQUIREMENTS 3.17 CONTROL ROOM HABITABILITY 4.17 CONTROL ROOM RABITABILITY Auplicability:

Apolicability:

Applies to the control room ventilation Applies to the periodic testing requirements system equipment necessary to maintain of systems required to maintain control room habitability.

. habitability.

Obiectives:

Objectives:

To assure the control room is habitable both To verify the operability of equipment under normal and accident conditions.

related to control room habitability.

Specification:

Soecification:

A. Control Room Ventilation System

'A.

Control Room Ventilation System

^='a na" chi check 4

1.

Except as specified in 3.17.A.2 and 1.

_ca"*

ontiol"r$om'1kspetatute.

3.17.A.3 below, both trains of the control room ventilation system shall be operable, whenever irradiated fuel is in the reactor vessel and reactor coolant temperature is greater than 212* F, or 4

movement of irradiated fuel

)

ep j g yg$

duringliesinthe seconda g containment, i

assem core alterations or activities having the potential for draining the reactor vessel.

2.a With one control room ventilation train inoperable, restore the inoperable train to operable status within 30 days.

2.b If 2.a is not met, then be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months following the 30 days and in cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.c If 2.a is not met during movement of irradiated fuel assemblies in the Nf_(_

s' M 9L7tR5 secondary containment, co e

\\g_ %

alterations or ng the potential for dra ning the reactor vessel then immediately place the operable control room ventilation train in operation or immediately suspend these activities.

229u 3.17/4.17 REV 139 8/25/94

Q

~

v bufpiCVM.wi 1 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization A..The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for the safe operation and maintenance of the plant.

During eriods when the Plant Manager is unavailable, this responsibility may be delegated to othy at1Tiesi_supeyvi snry personnel.

h'dT Sucerwsof.

Th ' L P :rintendent r, a designated individual during periods of absence from the control room or's office) shall be responsible for the control room command function.

B.

Offsite and Onsite Organizations Onsite and offsite organizations shall be established for plant operation and corporate x

management, respectively.

The onsite and offsite organizations shall include os tions for activities affecting plant safety.

p%gg g g l.

Lines of authority, responsibility and communication shall be establis%e4 D efined fo the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appr

late, in the form of organization charts, function descriptions of department :::p ra

b' ' " _e and relationships, and job descriptions for key personnel positions, or in equivalent forms of t

documentation. These requirements are documented in corporate and plant procedures, or the g

Updated Saf Report or the Operational Quality Assurance Plan.

g kresie uclear Generation shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the g

h staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety. This osition h s the r-c.r '"

r the Fire Protection Program.

(

Theindividuals(who@ttaintheop c evw twTsi.

T

\\

3.

ating sta an o carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall

/

have sufficient organizational freedom to ensure their independence from operating pressures.

6.1 232 REV tio 1/ Luj5 F

C.

Plant Staff 1.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.1.1.

2.

At least one licensed operator shall be in the control room when fuel is in the reactor.

3.

At least two licensed operators shall be present in the control room during cold startup, scheduled reactor shutdown, and during recoyery from eacto yr i

An individual qualified in radiation protection procedures sha 1 be(_r c'd

(.Dq Pl*N%T 0 t

_t when g

4.

fuel is in the reactor.

A ge.r-5.

All alterations of the reactor core shall be directly supervised by a license

(#

f 9[h#

Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling e-who has no other concurrent responsibil ties uring this

-t wenT N D

b ge.(fd-d h N e naintai en cite at all

,/

E-6.

A fire brigade of at least five member times.*

The fire brigade shall not include the three members of the shift

/

organization required for safe shutdown of the reactor from outside the contro room.

7.

The General Superintendent, Operations shall be formerly licensed as a Senior Reactor Operator or hold a current Senior Reactor Operator License.

s' 8.

At least one member of plant management holding a current Senior Reactor Operator Licejn

[

shall be assigned to the plant operations group on a long term basis (approximately t90 y

years).

This individual will not be assigned to a rotating shift.

D.

Each member of the unit staff shall meet or exceed the minimum qualificationsy ANSI N18.1-1971 for comparable positions, except for (1) the Superintendent Radictier Pr tes.se who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (3) the General Super!,d:nt, Operations who shall meet the requirement of ANSI N18.1-1971 except that NRC lic e requirements are as specified in A

e.e-Specification 6.1.C.7.

The training program shall e under the direction of a designated member ofyctherr. States Power management.

o

Fire Brigade composition may be less than the minimum requirgents for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of Fire Brigad members provided immediate action is taken to restore the Fire Brigade to within the minimum requirements.

6.1 233 REV 110 7/it/09

-- k (Su.pe ri swe wiesi

%..j i

A E.

.A training program for individuals serving in the fire brigade shall be maintained under the direction of a designated member of Northern States Power management. This program.shall meet the requirement of Section 27 of the NFPA Code - 1976 with the exception of trainini; scheduling.

Fire brigade training shall be scheduled as set forth in the training program.

F.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance

~pg3 Cy personnel.

Procedures shall include the following provisions:

Q

{ l,, M # cA ' bo#

1.

Adequate shif t coverage shall be maintained without reuCTne g o(*

O heavy use of overtime. The objective shall b have operating l

A po@p I

personnel work a normal he"r d,, '04.ar._ 3 while the plant is operating.

However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelinos shall be followed:

An individual should not be permitted to work more than a.

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time, b.

Overtime should be limited for all nuclear plant staff personnel so that total work time.does not exceed 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any

^48-hour period, not more than 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months in any seven day period, all excluding shift turnover cima.

Individuals should not be required to work more than 15 consecutive days without two consecutive days off.

A break of at least eight hours including shift turnover time c.

should be allowed between work periods.

d.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

6.I 234 REV i10 '1 '.,'? ?

6.2 Rev._

.nd Audit

%%4[.

Organizattensi units fer the review end audit of facility operations r.hn11 be c nstituted cnd have the responsibilities and authorities outlined below:

A. Safety Audit Committee (SAC)

The Safety Audit Committee provides the independent review of plant operation: from a nuclear safety standpoint. Audits of plant operation are conducted under the cognizance of the SAC.

l

1. Membership h

em

a. The SAC 'shall consist of at least five (S) persons.

g@

b. The SAC Chairman shall be-an NSP repres iv not ng line responsibility for operation of the plant, appointed re ent, Nuclear Generation. Other members shall be appointed by th Q re en, Nuc1 car Generation or by such other person as he may designate. W e Chairman shall appoint a Vice Chairman from the SAC membership to act in his absence.

c. No more than two members of the SAC shall be from groups holding line responsibility l

for operation of the plant.

d. A SAC member may appoint an alternate to serve in his absence, with concurrence of the 1

Chairman.

No more than one alternate shall serve on the SAC at any one time. The alternate member shall have voting rights.

2. Qualifications

a. % e SAC members should collectively have the capability required to review activities in the following areas: nuclear power plant operations, nuclear engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering, quality assurance practices, and other i

appropriate fields associated with the unique characteristics of the nuclear power l

plant.

8 n

f P

6.2 237

's u

Investigation of all ReportaNie Events and Events requiring Special Reports to the Comunission.

f.

g.

Revisions to the Facility Emergency Plan, the Facility Security Plan, and the Fire Protection Program.

h.

Operations Committee minutes to determine if matters considered by that Committee involve unreviewed or unresolved safety questions.

1.

Other nuclear safety mattera referred to the SAC by the Operations Committee, plant management or company l

i management.

J.

All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures, systems, or components.

1 k.

Reports of special inspections and audits conducted in accordance with specification 6.3.

1.

Changes to the Offsite Dose Calculation Manual (ODCH).

m.

Review of investigative reports of unplanned releases of radioactive material to the environs.

6.

Audit - The operation of the nuclear power plant shall be audited formally under the cognizance of the SAC to l

assure safe facility operation.

a.

Audits of selected aspects of plant operation, as delineated in ANSI N18.7-1976 as modified by the l

Operational quality Assurance Plan, shall be performed with a frequency commensurate with their nuclear safety significance and in a manner to assure that an audit of all nuclear safety-related activities is a

completed within a period of two years. The audits shall he performed in accordance with apnropriate written instructions and procedures.

t 2

b.

Audits of aspects of plant radioactive effluent treatment and radiological environmental monitoring shall he performed as follows:

j l.

Implementation of the Offsite Dose Calculation Manual and quality controls for effluent monitoring at l'

least once every two years.

i 2.

Implementation of the Process Control Program for solidification of radioactive vaste at least once every two years.

1 3.

The Radiological Environmental Monitoring Program and the results thereof, including quality controls, at 7

least once every year.

i c.

Periodic review of the audit program should be performed by the SAC at le t twice a year to assure its adequacy.

/

I I

d.

Written reports of the audits shall he reviewed by th residen uclear Generation, by the SAC at a scheduled meeting, and by members of Management sving responsib lity the areas audited.

bg 6.2 239 Amendment No. yg, f4 44. 59 FEB 161989 i

't

=

8-\\s-g b ufpkeYGewT1 CYNemW h 7.

Authority g

v

resident, uclear Generation.

The SAC shall be advisory to th e

havsT ham 6r 8.

Records Minutes shall by prepared and retained for al schedule meetings of the Safety Audit

@mmireauThAminutes shall be distribute ithin one month of the meeting to the

$fr, --V4ee rre uclear Generation, the C:::_ 1 ".:::;rr "n l::: "la..te, each member of the SAC, and others designated by the Chairman or Vice Chairman. There shall be a

~

formal approval of the minutes.

9.

Procedures A written charter for the SAC shall be prepared that contains:

a.

Subjects within the purview of the group.

b.

Responsibility and authority of the group.

c.

Mechanisms for convening meetings.

d.

Provisions of use of specialists or subgroups.

Authority to obtain access to the nuclear power plant operating record files e.

and operating personnel when assigned audit functions.

f.

Requirements for distributilon of reports and minutes prepared by the group to others in the NSP Organization.

6.2 240 REV 96 7/2/?'

,g, 6

hovergj B.

Operations Committee (OC) l.

Membership visors of the on-site supervisory staff.The Operations Committee shall consist of at least six (6) regular members d The Plant Manager shall serve as Chairman of the OC and shall appoint a regular member to act as Vice Chairman in his absence.

Alternates to the regular members shall be designated in writing by the Chairman, or Vice Chairman in the Chairman's absence, to serve on a temporary basis.

members of the Operations Committee at any one time.No more than two alternates shall participate as voting 2.

Meeting Frequency The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.

3.

Quorum a quorum shall include a majority of the membership, incl,uding the Chairman or Vice Chairman.

Responsibilities - The fol?.owing subjects shall by reviewed by the Operations Committee:

/p 4.

Proposed tests and experiments and their results.

a.

b.

Modifications to plant systems or equipment as described in the Updated Safety Analysis Report and having nuclear safety significance or which involve an unreviewed safety question as defined in 10 CFR 50.59.

Proposals which would effect permanent changes to normal and emergency operating c.

procedures and any other proposed changes or procedures that are determined by the Plant Manager to affect nuclear safety.

d.

Proposed changes to the Technical Specifications or operating license.

All repertad or suspected violations of Technical Sp'ecifications, operating lice.ase e.

requirements, administrative procedures, or operating procedures. Results of investi-gations, including evaluation and recommendations to prevent recurrence, will be reported, in writing, to the Canaral ".: nager "u:lcar Plant: and to the Chairman of the Safety Audit Committee.

~

p +%

yssueg Cmaarb

-N #

,,m m., _. _ _

~ ~

W

/JsP l

,/

I O

g l

f. Investigation of all Reportable Events and Events requiring Special Reports to the Commission.

g. Drills on emergency procedures (including plant evacuation) and adequacy of communication with off-site support groups.

h. All procedures required by these Technical Specifications, including impicmenting procedures,of the Emergency Plan and the Security Plan (except as exempted in Section 6.5.F), shall be reviewed with a frequency commensurate alth their safety significance but at an interval of not more than two years.

Committee.

1. Perform special reviews and investigations, as requested by the Safoty Audit J. Review of investigative reports of unplanned releases of radioactive material to the environs.

k. All changes to the Process Control Program (PCP) and the Of fsite Dose Calculation llanual (ODClf).

5. Authority The OC Shall be advisory to the Plant Pfanager. In the event of disagreement between the recommendations of the

>e the more conservative will be OC and the Plant flanager, the course determined by the Plant ?!anager '

.Presiden tiocicar Generation and the l o

followed. A written summary of the disagreement will be sent to th Chairman of the SAC for review.

Q p\\eM A

/J 6P

6. Records itinutes shall be recorded for all meetings of the 0 and shall identify all documentary material reviewed.

the OC, the Chairman and each member of the Safety Audit The minutes shall be distributed to each memberon and others designated by OC Chairman or Vice Chairman.

l Committee, th 44ee 44 siden uclear Cen N SP '

7. Procedures A written charter for the DC shall be prepared that contains:

in5 Responsibility and authority of the group, a.

Content and method of submission of presentations to the Operations Committee.

w b.

?

O 242 6.2

,8-W-%

w W

MN c.

Mechanism for scheduling meetings d.

Meeting agenda e.

Use of subcommittee

-t-f.

Review and approval, by members, of OC actions g.

Distribution of minutes 6.3 Special Inspections and Audits A.

An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified off-site Northern States Power Company personnel or an outside fire protection consultant, h

B.

An inspection and audit by an outside qualified fire protee on consultant shal be performed at intervals no greater than three years.

N '

e. hrc4le*T to r w emT't ON 6.4 Action to be Taken if a Safety Limit is Exceeded If a Safety Limit is exceeded, the reactor shall be shut d immediately. An immediate report shall be made to the Commission and to the cener:1 enager.!ueleer Plante, or his designated y

alternate in his absence. A complete' analysis of the circumstances leading up to and

[

resulting from the situation, together with recommendations by the operations Committee, shall also by prepared. This report shall by submitted to the Commission, to the Ocncral l'anagcr ikc. car I'lents and the Chairman of the Safety Audit Committee within 14 days of the occurrence.

Reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.

6.2 - 6.4 243 REV S'.

3/27/01

b T\\6 M by e.[\\ vgToddwh

)

b h&NOW 3.

Radiolocical

$cr Ac.c.$

1.

a.

A Radiation Protection Program, consistent with the requ ts of 10 CFR 20, shall be developed and followed. The Radiation Protection Pro shall consist of the following:

(1) A Radiation Protection Plan, which shall be a e lete definition of radiation protection policy and program (2) Procedures which implement the requirements of the R intion Protection Plan The Radiation Protection Plan and implementing proced es, with the exception of those non-safety related procedures governing work activi es exclusively applicable to or performed by health physics personnel, shall b reviewed by the Operations Committee and approved by a member of plant management des nated by the Plant Manager. Health physics procedures not reviewed by the Opers.tig Committee shall be reviewed and app-roved by the Superintendent, "cdicticr "rctcct cn b.

In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of f

10CFR20, each bigh radiation area in which the intensity of radiation is greater than 100

(

mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.1 Any individual or group of individuals permitted to enter such creas shall be provided with or accompanied by o e or more of the following:

T2 (1) A radiation monitoring device-wni ' ontinuously indicates the radiation dose rate in the area.

c(TeS $ W The, pegg,7 m h ue.

(2) A radiation monitoring device ontinuously integrates the radiation dose rate in the area and alarms when a prc c..: integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose m : 12 ::1 ir thc crce hoc bacr ET ac M liched and personnel have been made knowledgeable o t em.

/

Ib

[

Tg#

_ (3) An individual qualified in radiation protection procedures who ic :;cippcd with a radiation dose rate monitoring device. This individual ch_11 be responsible for providing positivedcontrol over the activities within the area and shall perform rdTCCTM periodic radiation surveillance at the frequency specified by th unit " cith "hg icist-

+A Radiation Work Permit.

The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr.

In addition doors shall be locked or attended,

[

to prevent unauthorized entry into these areas and the keys or key devices for locked doors shall be maintained under the administrative control of the Plant Manager.

l' Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from.

the Radiation Work Permit issuance requirement during the performance of their assigned

/

radiation protection duties, provided they comply with approved radiation protection procedures

[

for entry into high radiation areas. This footnote applies only to high radiation areas of 1000 mrem /hr or less.

6.5

^T

" " P

b 244a REV 120 3/2G/91-O C The o pp\\icnMQ,

~

~-

- - ~ ~

E.

Ot. site Dose Calculation Manual (ODCM)

The ODCM shall ba cpproved by the Commission prior to initial implementation. Changes to the ODCM shall satisfy the following requirements:

1.

Shall by submitted to the Commission with the Semi-Annual Radioactive Effluent release report for the period in which the change (s) were made effective. This submittal shall contain:

sufficiently detailed information to totally support the rationale for the change without a.

benefit of additionally or supplemental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a revision date, together with approp~riate analyses or evaluations justifying the change (s).

b.

a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c.

documentation of the fact that the change has been reviewed and found acceptable by the Operations Committee.

2.

Shall become effective upon review and acceptance by the Operations Committee.

F.

Security Procedures shall be developed to implement the requirements of the Security Plan and the Security Contingency Plan. These implementing procedures, with the exception of thc,se non-safety related procedures governing work activities exclusively applicable to or performed by security personnel, shall be reviewed by the Opc rations Committee and approved by a member of plant management designated by the Plant Manager.

Security procedures not reviewed by the Operations Committee shall be reviewed and approved by the Superintendent, Security eenddowo6ees.

j G.

Temporary Chances to Procedures Temporary changes to those procedures which are required to be reviewed by the Operations Committee described in A, B, C, D, E and F above, which do not change the intent of the original procedures may be made with the concurrence of two members of the unit management staff, at least one of whom holds a

((;

Senior operator License.

Such changes should be documented, reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager within one month. Temporary changes to health physics and security procedures not reviewed by the Operations Committee shall be reviewed by the E; xi-_:__ _, L Jial r ? ~ ~ ' - for health physics procedures and the Superintendent, Security * ' - 1: r forsecurityprocedures.[>

l 6.5 w

246b l

s genewg gweer.r<r e mm er,Pwhic,s 9,%

7' " "

vv v._

v. - %

_s

8-s-%

be\\d' B.

Records Retained for Plant Life (continued)

,~ 'M M.

- -- ; "' oder the pro,visioy of x,-

ar

'vered 11.

rds ironme al Qualif tion

~ _.

K Records of the service lives of all safety-related snubbers, including the date at which

(

c the service life commences and associated installation and maintenance records.

( %

./

6.6 247a REV 59 12/2?/?1

mg B.

Reportable Events The following actions shall be taken for Reportable Events:

The Commission shall be notified by a report submitted pursuant to the requirements of.

a.

Section 50.73 to 10 CFR Part 50 and, results b.

Each Reportable Event shall be reviewed by the Operations Conesittc a and of this review shall be submitted to the Safety Audit Committee and th siden N5 Nuclear Generation.

b q SMtW

~

r an 250-g 6.7 4

w

i:

l l

- Exhibit C

' Revised Technical Specification Pages

- Monticello Nuclear Generating Plant

('

Supplement 2 to License Amendment Request Dated August 15,1996 Exhibit C consists of revised Technical Specification pages that incorporate the proposed changes. The pages included in this exhibit are as listed below.

Page i, ii, iii, iv, v l

vi, vii, 22, 31, 61, 62 63a,69,72,82,89 99,102,125,126,164

[

188,190,198t,200 l

223,227b,227c,227d,227e 229b, 229c, 229ff, 229i, 229u 232,233,234,237,239 240,241,242,243,244a,246b 247a,250 l

i 1;

C-1

TABLE OF CONTENTS Eage 1.0 DEFINITIONS 1

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 and 2.3 Fuel Cladding integrity 6

2.1 Bases 10 2.3 Bases 14 2.2 and 2.4 Reactor Coolant System 21 2.2 Bases 22 l

2.4 Bases 24 3.0 ' LIMITING CONDITIONS FOR OPERATION AND 4.0 SURVEILLANCE REQUIREMENTS 4.0 Surveillance Requirements 25a j

4.0 Bases 25b 3.1 and 4.1 Reactor Protection System 26 3.1 Bases 35 4.1 Bases 42 l

3.2 and 4.2 Protective Instrumentation 45 A. Primary Containment Isolation Functions 45 B. Emergency Core Cooling Subsystems Actuation 46 C. Control Rod Block Actuation 46 i

D. Other Instrumentation 46a E. Reactor Building Ventilation isolation and Standby Gas Treatment System initiation 47 F. Recirculation Pump Trip initiation and Alternate Rod injection Initiation-48 G. Safeguards Bus Voltage Protection 48 H. Instrumentation for S/RV Low-Low Set Logic 48 1.

Instrumentation for Control Room Habitability Protection 48 3.2 Bases 64 4.2 Bases 72 3.3 and 4.3 Control Rod System 76 A. Reactivity Limitations 76 B. Control Rod Withdrawal 77 C. Scram insertion Times 81 D. Control Rod Accumulators 82 E. Reactivity Anomalies 83 F. Scram Discharge Volume 83a G. Required Action 83a 3.3 and 4.3 Bases 84 i

Amendment No. 30,37,45,65,

_ -. _. _. ~. _. _ _. _.

_ _ _ _ _ _. _. _.. _ _.. ~.

lI l

TABLE OF CONTENTS (Cont'd)

Ea9ft 3.4 and 4.4 '

Standby Liquid Control System 93 g

j.

A.JSystem Operation 93 B.J Boron Solution Requirements 95 C.-

96 3.4 and 4.'

Bases 99 4

I 3.5 and 4.5L Core and Containment / Spray Cooling Systems

101 l_

A. ECCS Systems 101 B. RHR Intertie Return Line isolation Valves 103 C Containment Spray / Cooling System 104

' D. RCIC 105 E. Cold Shutdown and Refueling Requirements 106 F. Recirculation System 107 3.5/4.5 Bases.

110-3.6 and 4.6 Primary System Boundary 121 A.- Reactor Coolant Heatup and Cooldown -

121 i

B. Reactor Vessel Temperature and Pressure 122 C.' _ Coolant Chemistry -

123 D.'. Coolant Leakage 126 E. _ Safety / Relief Valves 127 F. Deleted G. Jet Pumps -

128 H. Snubbers 129 3.6 and 4.6 Bases-

.145 l

3.7 and 4.7 Containment Systems 156-A. Primary Containment 156 B. Standby Gas Treatment System 166 C.. Secondary Containment 169 D. Primary Containment isolation Valves 170 E. Combustible Gas Control System 172 l

- 3.7 Bases 175 4.7 Bases 183 i

L jj 9/16/98 Amendment No. 9,35,47,74,77,79,400,102 4-

.-2

TABLE OF CONTENTS (Cont'd)

Eaga

- 3.8 and 4.8 Rad;oditive Effluents :

192 A. ' Liquid Effluents 192 B. Gaseous Effluents 197 C. Solid Radioactive Waste 198e t

D. Dose from All Uranium Fuel Cycle Sources 1981 3.8 and 4.8 Bases 198u 3.9 and 4.9 Auxiliary Electrical Systems 199 A. Operational Requirements for Startup 199 B. Operational Requirements for Continued Operation 200 1.

Transmission Lines 200 2.

Reserve Transformers 201 3.

Standby Diesel Generators 201 4.

Station Battery System 203 S.

24V Battery Systems 203 3.9 Bases 204 4.9 Bases 205 3.10 and 4.10 Refueling 206 A. Refueling Interlocks 206 8 Core Manitoring 207 i C. Fuel Storage Pool Water Level 207 D. Movement of Fuel 207 E. Extended Core and Control Rod Drive Maintenance 208 3.10 and 4.10 Bases 209

- 3.11 and 4.11 Reactor Fuel Assemblies 211 A. Average Planar Linear Heat Generation Rate 211 B. Linear Heet Generation Rate 212 C. Minimum Critical Power Ratio 213 3.11 Bases 216 4.11 Bases 218 l

3.12 and 4.12 Sealed Source Contamination 219 A. Contamination 219 B. Records 221 3.12 and 4.12 Bases 222 i

iii 9/16/98 Amendment No. 29,400,102

~. _... _ _ _

l :=

l t

l TABLE OF.COMTENTS (Cont'd)

Eage 3.13 and 4.13 Fire Detection Protection Systems 223 l,

"A. Fire Detection Instrumentation 223 B. Fire Suppression Water System 224 4

l-C.- Hose Stations 226 l

D. : Yard Hydrant Hose Houses 227-E. Sprinkler Systems :

227a F. Halon Systems

. 227b G. Penetration Fire Barriers

- 227b H. Alternate Shutdown System '

227c-3.13 Bases 228 4.13 Bases

- 228b l

3.14 and 4.14 Accident Monitoring Instrumentation 229a

- 3.14 and 4.14 Bases 229e 3.15 and 4.15 Inservice Inspection an'd Testing 229f-3.15 and 4.15 Bases 2299 3.16 and 4.16 Radiation Environmental Monitoring Program 229h

'A. ~ Sample Collection & Analysis 229h l

7 B. Land Use Census 229j C. Interlaboratory Comparison Program

'229k l

3.16 and 4.16 Bases 229t 3.17 and 4.17 Control Room Habitability 229u A. Control Room Ventilation System 229u B.. Control Room Emergency Filtration System 229v 3.17 Bases 229y 4.17 Bases 229z 5.0 DESIGN FEATURES 230 5.1 Site 230 5.2 Reactor -

230

' 5.3 Reactor Vessel 230

5.4 Containment 230 5.5 Fuel Storage 231 5.6 Seismic Designs 231 6.0 ADMINISTRATIVE CONTROLS-232

- 6.1 Organization 232 6.2 Review and Audit 237 6.3 SpecialInspection and Audits 243 l

6.4 Action to be taken if a Safety Limit is Exceeded 243 6.5 P.lant Operating Procedures 244 6.6 Plant Operating Records 246c l

6.7 Reporting Requirements 248 iv -

Amendment No. 45,37, -46, 61, 65, 6

-~-

m

- - -, ~ -,

l LIST OF FIGURES l-Figure No.

Page 3.4.1 Sodium Pentaborate Solution Volume-Concentration l

Requirements 97 l

3.4.2 Sodium Pentaborate Solution Temperature Requirements 98 I

l 3.6.1 Core Beltline Operating Limits Curve Adjustment

' vs. Fluence 133-3.6.2 Minimum Temperature vs. Pressure for Pressure Tests 134 i

3.6.3 Mir.imum Temperature vs. Pressure for Mechanical Heatup or l

Cooldown Without the Core Critical 135 l

'3.6.4 Minimum Temperature vs. Pressure for Core Operation 136 4.6.2 Chloride Stress Corrosion Test Results @ 500'F 137

' 3.7.1 Differential Pressure Decay Between the Drywell and Wetwell 19i p

3.8.1 Monticello Nuclear Generating Plant Site Boundary for Liquid Effluents 198g 3.8.2 Monticello Nuclear Generating Plant Site Boundary for Gaseous Effluents 198h l

I i

v Amendment No. 9,35,47,74,77,79,

l LIST OF TABLES Table No.

Page 3.1.1 Reactor Protection System (Scram) Instrument Requirements 28 l

l 4.1.1-Scram Instrument Functional Tests - Minimum Functional l

Test Frequencies for Safety instrumentation and Control Circuits 32 l

4.1.2 Scram instrument Calibration - Minimum Calibration Frequencies for Reactor Protection Instrument Channels 34 3.2.1 Instrumentation that Initiates Primary Containment Isolation Functions 49

[

3.2.2 Instrumentation that Initiates Emergency Core Cooling Systems 52 3.2.3 Instrumentation that Initiates Rod Block 56 l

l 3.2.4 Instrumentation that Initiates Reactor Building Ventilation isolation and Standby Gas Treatment System Initiation 59 3.2.5 Instrumentation That initiates a Recirculation Pump Trip and Alternate Rod Injection 60 3.2.6 Instrumentation for Safeguards Bus Degraded Voltage and Loss of Voltage Protection 60a l

3.2.7 Instrumentation for Safety / Relief Valve Low-Low Set Logic 60b 3.2.8 Other Instrumentation 60d 3.2.9 Instrumentation for Control Room Habitability Protection 60e

(

4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation 61 4.6.1 Snubber Visual Inspection Interval 132a l

j 3.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 1981 3.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198k i

4.8.1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 198m l

4.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 198n 4.8.3 Radioactive Liquid Waste Sampling and Analysis Program 198p l

4.8.4 Radioactive Gaseous Waste Sampling and Analysis Program 198s q

l i

4 i

l vi Amendment No. 37,39,-44,45,65,74,82,

l LIST OF TABLES (Cont'd)

' Table No.

Page 3.13.1 Safety Related Fire Detection Instruments 227d l

l 3.14.1 instrumentation for Accident Monitoring 229b 4.14.1 Minimum Test and Calibration Frequency for Accident j,

Monitoring Instrumentation 229d

4.16.1 Radiation Environmental Monitoring Program (REMP)'

Sample Collection and Analysis 229-1 4.16.2' REMP - Maximum Values for the Lower Limits of Detection (LLD) 229q l

4.16.3 REMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 229s 6.1.1 Minimum Shift Crew Composition 236 l-l vii Amendment No. 45,37A4,54,74

Bases 22:

The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The pressure safety limit of 1335 psig as measured in the vessel steam space is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value was derived from the design pressures of the reactor pressure vessel, coolant piping, and recirculation pump casing. The respective design pressures are 1250 psig at 575 F,1148 psig at 5620F, and 1380 psig l

at 575 F. The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes:

ASME Boiler and Pressure Vessel Code Section lil-A for the pressure vessel, ASME Boiler and Pressure Vessel Code Section ill-C for the recirculation pump casing, and the USAS Piping Code Section B31.1 for the reactor coolant system piping. The ASME Code permits pressure transients up to 10 percent over the vessel design pressure (110% x 1250 = 1375 psig) and the USAS Code permits pressure transients up to 20 percent over the piping design pressure (120% x 1148 = 1378 psig).

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig and temperature of 575oF; this is more than a factor of 1.5 below the yield strength of 42,300 psi at this temperature. At the pressure limit of 1375 psig, the general membrane stress increases to 29,400 psi, still safely below the yield strength.

The reactor coolant system piping provides a comparable margin of protection at the established pressure safety limit.

2.2 BASES 22 Amendment No. O,400a

1 Table 3.1.1 - Continued e.

The nigh drywell pressure scram functions in the Startup and Run modes when necessary'during purging for containment inerting or de-inerting only by closing the manual containment isolation valves. Verification of the bypass condition shall be noted in the control.

room log.

f.

One instrument channel for the functions indicated in the table to allow completion of surveillance testing, provided that:

1.

Redundant instrument channels in the same trip system are capable of initiating the automatic function and are demonstrated to i

be operable either immediately prior or immediately subsequent to applying the bypass.

2.

While the bypass is applied, surveillance testing shall proceed on a continuous basis and the remaining instrument channels initiating the same function are tested prior to any other. Upon completion of surveillance testing, the bypass is removed.'

i I

7

?

f t

3.1/4.1 31 I

Amendment No. O, 5

Table 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation Instrument Channel Test (3)

Calibration (3)

Sensor Check (3)

ECCS INSTRUMENTATION 1.

Reactor Low-Low Water Level Once/3 months (Note 5)

Every Operating Cycle - Transmitter Once/3 months -Trip Unit Once/12 hours l

2.

Drywell High Pressure Once/3 months Once/3 months None 3.

Reactor Low Pressure (Pump Start)

Once/3 months Once/3 months None 4.

Reactor Low Pressure (Valve Once/3 months Once/3 months None Permissive) 5.

Undervoltage Emergency Bus Refueling Outage Refueling Outage None 6.

Low Pressure Core Cooling Pumps Once/ month Once/3 months None Discharge Pressure Interlock 7.

Loss'of Auxiliary Power Refueling Outage Refueling Outage None 8.

Condensate Storage Tank Level Refueling Outage Refueling Outage None 9.

Reactor High Water Level Once/3 months (Note 5)

Every Operating Cycle - Transmitter Every 3 months - Trip Unit Once/12 hours l

ROD BLOCKS 1.

APRM Downscale Once/3 months (Note 5)

Once/3 months None 2.

APRM Flow Variable Once/3 months (Note 5)

Once/3 months None 3.

IRM Upscale Notes (2,5)

Note 2 Note 2 4.

IRM Downscale Notes (2,5)

Note 2 Note 2 5.

RBM Upscale Once/3 months (Note 5)

Once/3 months None 6.

RBM Downscale Once/3 rnonths (Note 5)

Once/3 months None 7.

SRM Upscale Notes (2,5)

Note 2 Note 2 8.

SRM Detector Not-Full-in Position Notes (2,9)

Note 2 None 9.

Scram Discharge Volume-High Level Once/3 months Refueling Outage None MAIN STEAM LINE (GROUP 1) ISOLATION 1.

Steam Tunnel High Temperature Refueling Outage Refueling Outage None 2.

Steam Line High Flow Once/3 months Once/3 Months Once/12 hours l

3.2/4.2 61 Amendment No. 2,40,37,39,63,66,31

k M.

i Table 4.2.1 Continued Minimum Test and Calibration Frequency for Core Cooling,.

i Rod Block and Isolation Instrumentation

}

instrument Channel Test (3)

Calibration (3)

Sensor Check (3) 3.

Steam Line Low Pressure -

Once/3 months Once/3 months None 4.

Reactor Low Low Water Level Once/3 months (Note 5)

Every Operating Cycle-Transmitter '

Once/12 hours -

l:

i Once/3 Months-Trip Unit l

CONTAINMENT ISOLATION (GROUPS 2 & 3) 1.

Reactor Low Water Level (Note 10) 2.

Drywell High Pressure (Note 10) 1 1

HPCI (GROUP 4) ISOLATION 1.

Steam Line High Flow Once/3 months Once/3 months None-r 2.

Steam Line High Temperature Once/3 months Once/3 months None l

RCIC (GROUP 5) ISOLATION 1.

Steam Line High Flow Once/3 months Once/3 months None 2.

Steam Line High Temperature Once/3 months Once/3 months None j

i REACTOR BUILDING VENTILATION & STANDBY GAS TREATMENT l

~

Reactor Low Low Water Level Once/3 months (Note 5)

Every Operating Cycle - Transmitter Once/12 hours l

1.

4 Once/3 months - Trip Unit -

i Drywell High Pressure (Note 10) 2.

3.

Radiation Monitors (Plenum).

Once/3 months Once/3 months Once/ day

(

4.

Radiation Monitors (Refueling Floor)

Once/3 months Once/3 months Note 4 i

'l RECIRCULATION PUMP TRIP AND ALTERNATE ROD INJECTION l

1.

Reactor High Pressure Once/3 months (Note 5)

Once/ Operating Cycle-Transmitter Once/ Day l

Once/3 Months-Trip Unit 2.

Reactor Low Low Water Level Once/3 months (Note 5)

Once/ Operating Cycle-Transmitter

' Once 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l

Once/3 Months-Trip Unit SHUIDOWN COOLING SUPPLY ISOLATION l

1.

Reactor Pressure Interlock Once/3 months Once/3 Months None f

i h

3.2/4.2 62' I

Amendment No. 74,81,83, Bt i

t

Table 4.2.1 Continued Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation NOTES:

(1)

(Deleted)

(2)

Calibrate prior to normal shutdown and start-up and thereafter check once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and test once per week until no longer l

required. Calibration of this instrument prior to normal shutdown means adjustment of channel trips so that they correspond, within acceptable range and accuracy, to a simulated signal injected into the instrument (not primary sensor). In addition, IRM gain adjustment will be performed, as necessary, in the APRM/lRM overlap region.

(3)

Functional tests, calibrations and sensor checks are not required when the systems are not required to be operable or are tripped. If tests are missed, they shall be performed prior to returning the systems to an operable status.

(4)

Whenever fuel handling is in process, a sensor ch' Sk shall be performed once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l (5)

A functional test of this instrument means the injection of a simulated signal into the instru nent (not primary sensor) to verify the proper instrument channel response alarm and/or initiating action.

(6)

(Deleted)

(7)

(Deleted)

(8)

Once/ shutdown if not tested during previous 3 month period.

(9)

Testing of the SRM Not-Full-in rod block is not required if the SRM detectors are secured in the full-in position.

(10) Uses contacts from scram system. Tested and calibrated in accordance with Tables 4.1.1 and 4.1.2.

3.2/4.2 63a Amendment No. 30,63,83

+

Bases 3.2 (Continuedh increases core voiding, a negative reactivity feedback. High pressure sensors initiate the pump trip in the event of an isolation transient. Low level sensors initiate the trip on loss of feedwater (and the resulting MSIV closure). The recirculation pump trip is only required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam which continues to be generated after reactor isolation in this unlikely postulated event, requiring the trip to be operable only when in the RUN mode is therefore conservative.

The ATWS high reactor pressure and low-low water level logic also initiates the Alternate Rod injection System. Two solenoid valves are installed in the scram air header upstream of the hydraulic control units. Each of the two trip systems energizes a valve to vent the header an causes rod insertion. This greatly reduces the long term consequences of an ATWS event.

Voltage sensing relays are provided on the safeguards bus to transfer the bus to an altemate source when a loss of vcitage condition or a degraded voltage condition is sensed. On loss of voltage this transfer occurs immediately. The transfer on degraded voltage has a time delay to prevent transfer during the starting of large loads. The degraded voltage setpoint corresponds to the minimum acceptable safeguards bus voltage for a steady state LOCA load that maintains adequate voltage at the 480V essential MCCS. An allowance for relay tolerance is included.

Safety / relief valve low-low set logic is provided to prevent any safety / relief valve from opening when there is an elevated water leg in the respective discharge line. A high water leg is formed immediately following valve closure due to the vacuum formed when steam condenses in the line. If the valve reopens before the discharge line vacuum breakers act to retum water level to normal, water clearing thrust loads on the discharge line may exceed their design limit. The logic reduces the opening setpoint and increases the blowdown range of three non-APRS valves following a scram. A 15-second interval between subsequent valve actuations is provided assuming one valve fails to 3.2 BASES 69 Amendment No. 30,34,45,400a

Bases 4.2:

The instrumentation in this section will be functionally tested and calibrated at regularly scheduled intervals. Although this,

instrumentation is not generally considered to be as important to plant safety as the Reactor Protection System, the same design reliability goals are applied. As discussed in Section 4.1 Bases, monthly or quarterly testing is generally specified unless the testing must be conducted during refueling outages. Quarterly calibration is specified unless the' calibration must be conducted during refueling outages. Where applicable, sensor checks are specified on a once/12 hours or once/ day basis.

l 4

4.2 BASES 72 NEXT PAGE IS 76 Amendment No. 63,81,400a i

i 1

3.0. LIMITING CONDITIONS FOR OPERATION.

4.0 SURVEILLANCE REQUIREMENTS l f

Any four rod group may contain a control rod which is valved out of service provided the above requirements and Specification 3.3.A are met.

D.

Control Rod Accumulators D. Control Rod Accumulators of the required Operable accumulator pressure and Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months check the status in the control room Control rod accumulators shall be operable in the Startup, Run, or Refuel modes except as provided below.

alarms.

1.

In the Startup or Run Mode, a rod accumulator may.

be inoperable provided that no other control rod in the nine-rod square array around this rod has a:

(a) Inoperable accumulator, or (b) Directional control valve electrically disarmed f

while in a non-fully inserted position.

If a control rod with an inoperable accumulator is inserted " full-in" and its directional control valves are i

electrically disarmed, it shall not be considered to have an inoperable accumulator.

I i

i 3.3/4.3 82 Amendment No. 5,44,43,54,63 f

L Bases 3.3/4.3 (Continued):

consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second

- assures that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions. One. operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism.

i

- C.

Scram insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the Safety Limit (T.S.2.1.A). This requires the negative reactivity insertion in any local region of the -

core and in the overall core to be equivalent to at least the scram reactivity curve used in the transient analysis. The required

~

average scram times for three control rods in all two by two arrays and the required average scram times for all control rods are based on inserting this amount of negative reactivity at the specified rate locally and in the overall core. Under these conditions, the CPR safety limit is never exceeded during any transient requiring control rod scram, and therefore MCPR remains above the Safety l.

Umit (T.S.2.1.A).

i 3.3/4.3 BASES '

89 Amendment No. 29,400a

Bases 3.4/4.4:

A.

The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of boron in the reactor core in less than 125 minutes sufficient to bring the reactor from full power to a 3% delta k subcritical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 25% boron concentration margin to allow for leakage and imperfect mixing.

[

The time requirement (125 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.

The ATNS Rule (10 CFR 50.62) requires the addition of a new design requirement to the generic SLC System design basis.

i Changes to flow rate, solution concentration or boron enrichment to meet the ATWS Rule do not invalidate the original system design basis. Paragraph (c)(4) of 10 CFR 50.62 states that:

"Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution' (natural boron enrichment).

The described minimum system parameters (equivalent to 24 gpm,10.7% concentration and 55 atom percent Boron-10 enrichment) will ensure an equivalent injection capability that meets the ATWS rule requirement.

Boron enrichment concentration, solution temperature, and volume (including check of tank heater and pipe heat tracing system) are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. A reliability analysis indicates that the plant can be operated safely in this manner for ten days. For additional margin, the allowable out of service time has been reduced to seven days.

The only practical time to test the standby liquid control system is during a refueling outage and by initiation from local stations.

Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the replacement charges for the tested system are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control room.

The relief valves in the standby liquid control system protect the system piping and positive displacement pumps which are nominally designed for 1500 psi from overpressure. The pressure relief valves discharge back to the standby liquid control solution tank.

3.4/4.4 BASES 99 Amendment No. 56,57,77,400a

Y 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.

One of the following conditions of inoperability may 4.

Perform the fo!!owing tests:

exist for the period specified:

Jigm Frecuency a.

One Core Spray subsystem may be inoperable for 7 days, or Motor Operated Pursuant to Valve Operability Specification b.

One RHR pump may be inoperable for 30 days, 4.15.B or ADS Valve Each Operating c.

One low pressure pump or valve (Core Spray or Operability Cycle RHR) may be inoperable with an ADS valve Note: Safety / relief valve operability is verified by inoperable for 7 days, or cycling the valve and observing a compensating d.

One of the two LPCIinjection paths may be change in turbine bypass or control valve position.

l inoperable for 7 days, or ADS Inhibit Each Operating Two RHR pumps may be inoperable for 7 days, Switch Operability Cycle e.

or Perform a simulated Each Operating automade actuadon test Cycle f.

Both of the LPCI injection paths may be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or (including HPCI transfer to the suppression pool and g.

HPCI may be inoperable for 14 days, provided automatic restart on RCIC is operable, or subsequent Icw reactor water level) h.

One ADS valve may be inoperable for 14 days, or 5.

Perform the following test on the Core Spray Ap Instrumentation i.

Two or more ADS valves may be inoperable for 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s-Check Once/ day 4.

If the requirements or conditions of 3.5.A.1,2 or 3 Test Once/ month

^

cannot be met, an orderly shutdown of the reactor Calibrate Once/3 months shall be initiated and the reactor shall be placed in a condition in which the affected equipment is not required to be operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.5/4.5 102 Amendment No. 7J,79 1

..-___ ___-_ _ - _ _ - _ _ _ _ _ = _ _ _

t' 3.0 LIMIT!NG CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 2.

(a) The reactor coolant water shall not exceed the 2.

During startup and at steaming rates below 100,000 following limits with steaming rates less than pounds per hour, a sample of reactor coolant shall 100,000 pounds per hour except as specified in be taken every four hours and analyzed for 3.6.C.2.b.

conductivity and chloride content.

Conductivity 5 mho/cm Chloride ion 0.1 ppm (b) For reactor startups the maximum value for conductivity shall not exceed 10 mho/cm and the maximum value for chloride ion concentration shall not exceed 0.1 ppm for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after placing the reactor in the power operating condition.

t L

3.

Except as specified in 3.6.C.2.b above, the reactor 3.(a) With steaming rates greater than or equal to coolant water shall not exceed the following limits 100,000 lbs. per hour, a reactor coolant sample l

with steaming rates greater than or equal to shall be taken at least every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and when the 100,000 lbs. per hour.

continuous conductivity monitors indicate abnormal conductivity (other than short-term spikes) and Conductivity 5 mho/cm analyzed for conductivity and chloride ion content.

Chloride ion 0.5 ppm

@)

en e condnuous con &ctMy manhods 4.

If Specifications 3.6.C.1 through 3.6.C.3 are not in perable, during power operation, a reactor met, an orderly shutdown shall be initiated and the coolant sample should be taken once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l

reactor shall be. the cold shutdown condition and analyzed for conductivity and chloride ion in within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

content.

i 3.6/4.6 125 Amendment No. O p

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS D.

Coolant Leakage D.

Coolant Leakage 1.

Any time irradiated fuel is in the reactor vessel and 1.

Any time irradiated fuel is in the reactor vessel and coolant temperature is above 212 F, reactor coolant coolant temperature is above 212oF, the following system leakage, based on sump monitoring, shall surveillance program shall be carried out:

be limited to:

a.

Unidentified and Identified Leakage rates shall a.

5 gpm Unidentified Leakage be recorded once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months using primary l

b.

2 gpm increase in Unidentified Leakage within containment floor and equipment drain sump any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period monitoring equipment.

t c.

20 gpm Identified Leakage 2.

The reactor coolant system leakage detection d.

no pressure boundary leakage systems shall be demonstrated OPERABLE by:

2.

With reactor coolant system leakage greater than a.

Primary containment atmosphere particulate 3.6.D.1.a or 3.6.D.1.c above, reduce the leakage monitoring systems-performance of a sensor rate to within acceptable limits within four hours or check once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a channel functional l

initiate an orderly shutdown of the reactor and test at least monthly and a channel calibration reduce reactor water temperature to less than at least once per cycle.

j 212 F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

Primary containment sump leakage 3.

With an increase in Unidentified Leakage in excess measurement system-performance of a sensor of the rate specified in 3.6.D.1.b, identify the source check once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a channel l

of increased leakage within four hours or initiate an calibration test at least once per cycle.

orderly shutdown of the reactor and reduce reactor water temperature to less than 212oF within 24 i

hours.

4.

If any Pressure Boundary Leakage is detected when the corrective actions outlined in 3.6.D.2 and i

3.6.D.3 above are taken, initiate an orderly i

shutdown of the reactor and reduce reactor water temperature to less than 212oF within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.6/4.6 126 Amendment No. 45,47,87

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 7

i 4.

Pressure Suppression Chamber-Drywell Vacuum 4.

Pressure Suppression Chamber-Drywell Vacuum Breakers Breakers a.

When primary containment integrity is required, a.

Operability and full closure of the l

611 eight drywell-suppression chamber vacuum drywell-suppression chamber vacuum breakers breakers shall be operable and positioned in shall be verified by performance of the the clos ad position as indicated by the position following:

indicatic a system, except during testing and except as specified in 3.7.A.4.b through (1) Monthly each operable drywell-suppression.

3.7.A.4.d below.

chamber vacuum breaker shall be exercised through an opening-closing b.

Any drywell-suppression chamber vacuum cycle.

breaker may be nonfully closed as indicated by the position indication and alarm system

- (2) Once each operating cycle, drywell to provided that drywell to suppression chamber suppression chamber leakage shall be differential pressure decay does not exceed demonstrated to be less than that that shown on Figure 3.7.1 equivalent to a one-inch diameter orifice and each vacuum breaker shall be visually c.

Up to two drywell-suppression chamber vacuum inspected. (Containment access required) breakers may be inoperable provided that: (1) the vacuum breakers are determined to be fully (3) Once each operating cycle, vacuum closed and at least one posit;on alarm circuit is breaker position indication and alarm operable or (2) the vacuum breaker is secured systems shall be calibrated and functionally in the closed position or replaced by a blank tested. (Containment access required) flange.

(4) Jn> e each operating cycle, the vacuum weakers shall be tested to determine that d.

Drywell-suppression chamber vacuum breakers the force required to oper; each valve from may be cycled, one at a time, during 7

fully closed to fully open does not exceed containment inerting and deinerting operations that equivalent to 0.5 psi acting on the to assist in purging air or nitrogen from the suppression chamber face of the valve suppression chamber vent header.

disc. (Conta,mment access required.)

l 3.7/4.7 164 l

Amendment No. 8,36,80 i

Bases 4.7 (Continued):

B.

Standby Gas Treatment System, and C. Secondary Containment initiating reactor building isolation and operation of the standby gas treatment system to maintain the design negative pressure within the secondary containment provides an adequate test of the reactor building isolation valves and the standby gas treatment system. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment sysum operational capability. Secondary Containment Capability Test data obtained under non-calm conditions is to be ext apolated to calm wind conditions using information provided in "Surnmary Technical Report to the United States Atomic Energy Commission, Directorate of Licensing, on Secondary Containment Leak Rate Test", submitted by letter dated July 23,1973, and as described in NSP letter to the NRC dated August 18,1995, with subject," Revision 2 to License Amendment Request Dated June 8,1994, Standby Gas Treatment and Secondary Containment Technical Specifications."

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Standby gas treatment system inplace testing procedures will be established utilizing applicable sections of ANSI N510-1989 standard as a procedural guideline only. If painting, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis should be performed as required for operational use. Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52 Revision 2 (March 1978) except testing should be IAW D3803-1989. The charcoal l

adsorber efficiency test procedures will allow for the removal of a representative sample. The 30 C,95% relative humidity test per ASTM D 3803-89 is the test method to establish the methyl iodine removal efficiency of adsorbent. The sample will be at least two inches in diameter and a length equal to the thickness of the bed. If the iodine removal efficiency test results are unacceptable, all adsorbent in the system will be replaced. High efficiency particulate filters are installed before and after the charcoal filters to prevent clogging of the carbon adsorbers and to minimize potential release of particulates to the environment.

An efficiency of 99% is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by inplace testing with DOP as the testing medium. Any HEPA filters found defective will be replaced with filters qualified pursuant to regulatory guide position C.3.d of Regulatory Guide 1.52 Revision 2 (March 1978). Once per operating cycle demonstration of HEPA filter pressure drop, operability of inline heaters at rated power, automatic initiation of each standby gas treatment system circuit, and leakage tests after maintenance or testing which could affect leakage, is necessary to assure system performance capability.

4.7 BASES 188 Amendment No. 94,400a

Bases 4.7 (Continued):

The containment is penetrated by a large number of small diameter instrument lines. A program for the periodic testing (see Specification 4.7.D) and examination of the valves in these lines has been developed and a report covering this program was.

submitted to the AEC on July 27,1973.

l' The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of re!iability.

E.

Combustible Gas Control System The Combustible Gas Control System (CGCS) is functionally tested once every six months to ensure that the recombiner trains will be available if required. In addition, calibration and maintenance of essential components is specified once each operating cycle.

a t

1 4.7 BASES 190 Amendment No. 36,400a

i

?

TABLE 4.8.4 - RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (Continued)

(Page 2 of 2) f Notes:

a.

The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal. Note (a) of Table 4.8.3 is applicable.

b.

Grab samples taken at the discharge of the plant stack and reactor building vent are generally below minimum detectable levels for l

most nuclides with existing analytical equipment. For this reason, isotopic analysis data, corrected for holdup time, for samples taken at the steam jet air ejector may be used to calculate noble gas ratie.

c.

Whenever the steady state radiciodine concentration is greater than 10 percent of the limit of Specification 3.6.C.1, daily sampling of reactor coolant for radioactive iodines of I-131 through I-135 is required. Whenever a change of 25% or more in calculated Dose Equivalent 1-131 is detected under these conditions, the iodine and particulate collection devices for all release points sha!! be l

removed and analyzed daily until it is shown that a pattern exists which can be used to predict the release rate. Sampling may then revert to weekly. When samples collected for one day are analyzed, the corresponding LLD's may be increased by a factor of 10.

Samples shall be analyzed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after removal.

d.

To be representative of the average quantities and concentrations of radioactive materials in particulate form in gaseous effluents, samples should be collected in proportion to the rate of flow of the effluent streams.

e.

The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, l

Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported.

}

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

I f.

Nuclides which are below the LLD for the analyses shall be reported as "less than" the LLD of the nuclide and should not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations. When unusual circumstances result in LLD's higher than reported, the reasons shall be documented in the semiannual effluent report.

g.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period sampled.

h.

H3 analysis shall not be required prior to purging if the limits of 3.8.B.1 are satisfied for other nuclides. However, the H analysis shall i

3 be completed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after sampling.

i.

In lieu of grab samples, continuous monitoring with bi-weekly analysis using silica-gel samplers may be provided.

l 3.8/4.8 198t Amendment No. 45,90

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 2.

Both diesel generators are operable and capable of feeding their designated 4160 volt buses.

3.(a) 4160V Buses #15 and #16 are energized.

(b) 480V Load Centers #103 and #104 are energized.

4.

All station 24/48,125, and 250 volt batteries are charged and in service, and associated battery chargers are operable.

B.

When the mode switch is in Run, the availability of electric power shall be as specified in 3.9.A, except as l

specified in 3.9.B or the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1.

Transmission Lines From and after the date that incoming power is available from only one line, reactor operation is permissible only during the succeeding seven days unless an additional line is sooner placed in service providing both the emergency diesel generators are operable.

I 3.9/4.9 200 Amendment No. 54

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.13 FIRE DETECTION AND PROTEC90N SYSTEMS 4.13 FIRE DETECTION AND PROTECTION SYSTEMS Acolicability:

Acolicability:

Applies to instrumentation and plant systems used for fire applies to the periodic testing of instrumentation and plant detection and protection of the nuclear safety-related systems used for fire detection and protection of the nuclear l

structures, systems, and components of the plant.

safety related structures, systems, and components.

Obiective:

Objective:

To insure that the structures, systems, and components of To verify the operability of instrumentation and plant systems the plant important to nuclear safety are protected from fire used for fire detection and protection of nuclear safety damage.

related structures, systems, and components.

Soecification:

A.

Fire Detection Instrumentation A.

Fire Detection Instrumentation 1.

Except as specifieo below, the minimum fire 1.

Fire detection instrumentation in each of the zones detection instrumentation for each fire detection in Table 3.13.1 shall be demonstrated operable zone shown in Table 3.13.1 shall be operable every six months by performance of functional tests.

whenever equipment in that fire detection zone is required to be operable.

2.

Alarm circuitry associated with the fire detector l

instruments in each of the zones in Table 3.13.1 2.

If specification 3.13.A.1 cannot be met, within one shall be demonstrated operable every six months.

hour establish a fire watch patrol to inspect the zone (s) with inoperable instruments once per hour

(+ 25%). Restore the minimum number of instruments to operable status within 14 days or submit a special report to the Commission within 30 days outlining the cause of the inoperability and the plans and schedule for restoring the instruments to operable status.

3.13/4.13 223 Amendment No. 7,46

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS F.

tialon Systems F.

Halon Systems 1.

The cable spreading room Halon system shall be 1.

The cable spreading room Halon system shall be operable with the storage tanks having demonstrated operable as fo!!ows:

l at least 95% of full charge weight and 90% of full a.

Each valve (manual, power operated, or charge pressure.

automatic) in the flow path that is not electrically superv sed, locked, sealed or otherwise 2.

If specification 3.13.F.1 cannot be met, within one secured in position, shall be verified to be,n,ts i i hour establish a continuous fire watch with backup rrect position every month.

fire suppression equipment in the cable spreading room. Restore the system to operable status within b.

Verify Halon storage tank weight and pressure 14 days or submit a special report to the every six months.

Cornmission within 30 days outlining the cause of Peh a gsm Wiod M eg M the inoperability and the plans and schedule for restoring the system to operable status.

m nths which includes verifying the system, including associated ventilation dampers, G.

Penetration Fire Barriers actuates manually and automatically, upon re eipt of a test signal.

1.

All penetration fire barriers in fire area boundaries shall be operable whenever safe shutdown d.

Perform an air flow test every 3 years through equipment in that fire area is required to be headers and nozzles to assure no blockage.

Operable.

e.

visuatiy examine headers and nozzies every 18 2.

If Specification 3.13.G.1 cannot be met, a months. An air flow test shall be performed continuous fire watch shall be established on at upon evidence of obstructions of any Halon least one side of the affected penetration (s) within system nozzle.

l one hour or verify the operability of fire detectors on G.

Penetration Fire Barriers at least one side of the non-functional fire barrier l

and establish an hourly (+ 25%) fire watch patrol.

1.

A visual inspection of penetration fire barriers in fire Restore the inoperable penetration fire barriers to area boundaries protecting safe shutdown Operable status within 14 days or submit a special equipment shall be conducted every 18 months.

report to the Commission within 30 days outlining the cause of the inoperability and the plans and 2.

Following repair or maintenance of a penetration fire schedule for restoring the barriers to Operable barrier a visualinspection of the seal shall be status.

conducted.

3.13/4.13 227b Amendment No. 7,46,61

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVElLLANCE REQUIREMENTS H.

Alternate Shutdown System H.

Altemate Shutdown System 1.

The system controls on the ASDS panel shall be 1.

Switches on the alternate shutdown system panel operable whenever that system / component is shall be functionally tested once per operating required to be operable.

cycle.-

2.

If system controls required to be operable by Specification 3.13.H.1 are made or found 2.

The alternate shutdown system panel master inoperable, restore the inoperable system control to -

transfer switch shall be verified to alarm in the operable within 7 days, or perform one of the control room when unlocked once per operating following; cycle.

a.

Provide equivalent shutdown capability and within 60 days restore the inoperable system controls to operable; or b.

Establish a continuous fire watch in the cable spreading room and the back-panel area of the control room and within 60 days restore the inoperable system controls to operable; or c.

Verify the operability of the fire detectors in the cable spreading room and the back-panel area of the control room and establish a hourly fire watch patrol and within 60 days restore the inoperable system controls to operable; or d.

Place the reactor in a condition where the 1

systems for whien the system controls at the ASDS are inoperable are not required to be operable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

The altemate shutdown system panel master transfer switch shall be locked in the normal position except when in use, being tested or being i

maintained.

3.13/4.13 227c f

Amendment No. 47, 61

TABLE 3.13.1 SAFETY RELATED FIRE DETECTION INSTRUMENTS Fire Detection Location Minimum Instruments Operable l

Zone Heat Flame Smoke 1A

B" RHR Room

.3 iB "A" RHR Room 3

1C RCIC Room 3

1E HPCIRoom 2

1F Reactor Building-Torus Compartment 11 2A Reactor Bldg. 935' elev - TIP Drive Area 1

2B Reactor Bldg. 935' elev - CRD HCU Area East 10 2C Reactor Bldg. 935' elev - CRD HCU Area West 11 2G/2H Reactor Bldg. 935' - LPCI Injection Valve Area 1

l 3B Reactor Bldg. 962' elev - SBLC Area 2

3C Reactor Bldg. 962' elev - South 5

3D Reactor Bldg. 962' elev - RBCCW Pump Area 4

4A Reactor Bldg. 985' elev - South 4

4B Reactor Bldg. 985' elev - RBCCW Hx Area 5

4D SBGT System Room 2

5A Reactor Bldg.1001' elev - South 7

5B Reactor Bldg.1001' elev - North 3

SC Reactor Bldg. - Fuel Pool Cooling Pump Area 1

6 Reactor Building 1027' elev 5

7A Battery Room 1

7B Battery Room 1

7C Battery Room 1

8 Cable Spreading Room 7

227d 3.13/4.13 Amendment No. 61

TABLE 3.13.1 (Continued)

SAFETY RELATED FIRE DETECTION INSTRUMENTS Fire Detection Location Minimum Instruments Operable l

~

' Zone Heat -

Flame Smoke P

12A Turbine Bldg. - 911' - 4.16 KV Switchgear 3

13C Turbine Bldg. - 911' elev - MCC 133 Area 1

-l 14A Turbine Bldg. - 931' - 4.16 KV Switchgear 2

15A/15C

12 DG Room & Day Tank Room 3

15B/15D

11 DG Room & Day Tank Room 3

16 Turbine Bldg. 931' elev - Cable Corridor 3

17 Turbine Bldg. 941' elev - Cable Corridor 3

19A Turbine Bldg. 931' elev - Water Treatment Area 5

19B Turbine Bldg. 931' elev - MCC 142-143 Area 1

19C Turbine Bldg. 931' elev - FW Pipe Chase 1

20 Heating Boiler Room 1

23A Intake Structure Pump Roca 3

31A 1st Floor - Reactor Building Addition - Division i 3

31B 1st Floor - Reactor Building Addition - Division 11 15 32A 2nd Floor - Reactor Building Addition - Division 1 6

32B 2nd Floor - Reactor Building Addition - Division 11 4

33 3rd Floor - Reactor Building Addition 5

227e 3.13/4.13 Amendment No. 61

Table 3.14.1 Instrumentation for Accident Monitoring Function Total No. of Minimum No. of Required Instrument Channels Operable Channels Conditions

Reactor Vessel Fuel Zone Water Level 2

1 A, B Safety / Relief Valve Position 2

1 A, C -

(One Channel Pressure Switch and One Channel Thermocouple Position Indication per Valve)

Drywell Wide Range Pressure 2

1 A, B Suppression Pool Wide Range Level 2

1 A, 8 Suppression Pool Temperature 2

1 A, D Drywell High Range Radiation 2

1 A, D Drywell and Suppression Pool 2

1 A, B Hydrogen and Oxygen Monitor Offgas Stack Wide Range Radiation 2

1 A, D Reactor Bldg Vent Wide Range Radiation 2

1 A, D

Required Conditions A.

When the number of channels made or found to be inoperable is such that the number of operable channels is less than the total number of channels, either restore the inoperable channels to operable status within seven days, or prepare and submit a special report to the Commie :icn pursuant to Technica Specification 6.7.D within the next 30 days outlining the action taken, the cause of l

the inoperability, m - ' u plans and schedule for restoring the system to operable status.

3.14/4.14 229b Amendment No. 2,37,63

Table 3.14.1 (Continued)

Instrumentation for Accident Monitoring

-

Required Conditions (continued)

B.

When the number of channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, the minimum number of channels shall be restored to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C.

When the number of channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, the torus temperature shall be monitored once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (+25%) to observe any l

unexplained temperature increase which might be indicative of an open SRV; the minimum number of channels shall be restored to operable status within 30 days or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D.

When the number of channels made or found to be inoperable is such that the number of operable channels is less than the' r

minimum number of operable channels shown, initiate the preplanned altemate method of monitoring the appropriate parameters in addition to submitting the report required in (A) above.

3.14/4.14 229c Amendment No. 2,37,63

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS B.

Inservice Testing 1.

Inservice Testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with the requirements for ASME Code Class 1,2 and 3 pumps and valves, respectively, contained in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) except where relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55(a)(g)(6)(i), or where alternate testing is justified in accordance with Generic Letter 89-04.

2.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

3.15/4.15 229ff Amendment No. 6,37,72,77

I P

P 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.

Deviations are permitted from the required sampling schedule if samples are unobtainable due to hazardous conditions, seasonable unavailability, or to malfunction of automatic sampling equipment. If the latter occurs, every effort shall be made to complete corrective action prior to the end of the next sampling period.

4.

With the level of radioactivity in an environmental

[

sampling medium exceeding the reporting levels of Table 4.16.3 when averaged over any calendar quarter, submit a special report to the Commission within 30 days from the end of the affected calendar quarter pursuant to Specification 6.7.C.2. When l

more than one of the radionuclides in Table 4.16.3 l

are detected in the sampling medium, this report i

shall be submitted if:

concentration (1), concentration (2) + - >10 limit level (1) limit level (2) t When radionuclides other than those in Table 4.16.3 h

are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.8.A.2, 3.8.B.2, or 3.8.B.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiation Environmental Monitoring Report.

3.16/4.16 229i Amendment No. 37,39,46 l

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.17 CONTROL ROOM HABITABILITY 4.17 CONTROL ROOM HABITABILITY Acolicability:

Acolicability:

Applies to the control room ventilation system equipment Applies to the periodic testing requirements of systems necessary to maintain habitability.

required to maintain control room habitability.

Obiective-Obiective:

To assure the control rcom is habitable both under normal and accident conditions.

To verify the operability of equipment related to control room Soecification:

A.

Control Room Ventilation System 1.

Except as specified in 3.17.A.2 and 3.17.A.3 below, both A.

Control Room Ventilation System trains of the control room ventilation system shall be operable, whenever irradiated fuel is in the reactor 1.

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months check control room l

vessel and reactor coolant temperature is greater than temperature.

212 F, or during movement of irradiated fuel assemblies in the secondary containment, core alterations or activities having the potential for draining the reactor vessel.

2.a With one control room ventilation train inoperable, restore the inoperable train to operable status within 30 days.

2.b If 2.a is not met, then be in hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months following the 30 days and in cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.c if 2.a is not met during movement of irradiated fuel assemblies in the secondary containment, core l

alterations or activities having the potential for draining the reactor vessel then immediately place the operable control room ventilation train in operation or immediately suspend these activities.

L 3.17/4.17 229u Arnendment No. 65,89 l

6.0 ADMINISTRATIVE CONTROLS 6.1 Organization A. - The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for the safe operation and maintenance of the plant. During periods when the Plant Manager is unavailable, this responsibility may be delegated to other qualified supervisory personnel.

The Shift Supervisor (or, a designated individual during periods of absence from the control room and shift supervisor's l

office) shall be responsible for the control room command function.

B.

Offsite and Onsite Organizations Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include positions for activities affecting plant safety.

1.

Lines of authority, responsibility and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, function descriptions of department responsib:lities and relationships, and job descriptions for key personnel positions, or in equivalent forms of l

documentation. These requirements are documented in corporate and plant procedures, or the Updated Safety.

Analysis Report or the Operational Quality Assurance Plan.

2.

The President, NSP Nuclear Generation shall have corporate responsibility for overall plant nuclear safety and shall l

take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety. This position has the responsibility for the Fire Protection l

Program.

3.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

l 6.1 232 Amendment No. 7,64,68

C.

Plant Staff 1.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.1.1.

2.

At least one licensed operator shall be in the control room when fuel is in the reactor.

3.

At least two licensed operators shall be present in the control room during cold startup, scheduled reactor shutdown, and during recovery from reactor trips.

4.

An individual qualified in radiation protection procedures shall be onsite when fuel is in the reactor.

l 5.

All alterations of the reactor core shall be directly supervised by a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

6.

A fire brigade of at least five members shall be maintained onsite at all times.* The fire brigade shall not include the l

three members of the shift organization required for safe shutdown of the reactor from outside the control room.

7.

The General Superintendent, Operations shall be formerly licensed as a Senior Reactor Operator or hold a current Senior Reactor Operator License.

8.

At least one member of plant management holding a current Senior Reactor Operator License shall be assigned to the plant operations group on a long term basis (approximately two years). This individual will not be assigned to a rotating shift.

D.

Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the General Superintendent Radiatien Services who shall meet or exceed the qualifications of l

Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (3) the General Superintendent, Operations who shall meet the requirement of ANSI l

N18.1 -1971 except that NRC license requirements are as specified in Specification 6.1.C.7. The training program shall be under the direction of a designated member of Northern States Power management.

l Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade to witWDthe minimum requirements.

6.1 233 Amendment No. 46,37,68

E.

A training program forindividuals serving in the fire brigade shall be maintained under the direction of a designated member of Northem States Power management. This program shall meet the requirement of Section 27 of the NFPA Code - 1976 with the exception of training scheduling. Fire brigade training shall be scheduled as set forth in the training program.

F.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel. Procedures shallinclude the following provisions:

1.

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8 or 12-hour day, nominal 40-hour week while the plant is operating. However, in l

the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

a.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift tumover time.

b.

Overtime should be limited for all nuclear plant staff personnel so that total work time does not exceed 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, not more than 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months in any seven day period, all excluding shift tumover time. Individuals should not be required to work more than 15 consecutive days without two consecutive days off.

c.

A break of at least eight hours including shift tumover time should be allowed between work periods.

d.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

6.1 234 Amendment No. 3,46,46,68

6.2 Review and Audit Organizational units for the review and audit of facility operations shall be constituted and have the responsibilities and authorities outlined below:

A.

Safety Audit Committee (SAC)

The Safety Audit Committee provides the independent review of plant operations from a nuclear safety standpoint. Audits of plant operation are conducted under the cognizance of the SAC.

1.

Membership a.

The SAC shall consist of at least five (5) persons.

b.

The SAC Chairman shall be an NSP representative, not having line responsibility for operation of the plant, appointed by the President, NSP Nuclear Generation. Other members shall be appointed by the President, NSP Nuclear l

Generation or by such other person as he may designate. The Chairman shall appoint a Vice Chairman from the SAC membership to act in his absence.

c.

No more than two members of the SAC shall be from groups holding line responsibility for operation of the plant.

d.

A SAC member may appoint an alternate to serve in his absence, with concurrence of the Chairman. No more than one alternate shall serve on the SAC at any one time. The alternate member shall have voting rights.

2.

Qualifications a.

The SAC members should collectively have the capability required to review activities in the following areas: nuclear power plant operations, nuclear engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering, quality assurance practices, and other appropriate fields associated with the unique characteristics of the nuclear power plant.

6.2 237 Amendment No. 3,46,46

f.

Investigation of all Reportable Events and Events requinng Special Reports to the Commission.

g.

Revisions to the Facility Emergency Plan, the Facility Security Plan, and the Fire Protection Program.

h.

Operations Committee minutes to determine if matters considered by that Committee involve unreviewed or unresolved safety questions.

i.

Other nuclear safety matters referred to the SAC by

'e ~'erations Committee, plant management or company management.

j.

All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures, systems, or components.

k.

Reports of special inspections and audits conducted in accordance with specification 6.3.

1.

Changes to the Offsite Dose Calculation Manual (ODCM).

m.

Review of investigative reports of unplanned releases of radioactive material to the environs.

6.

Audit - The operation of the nuclear power plant shall be audited formally under the cognizance of the SAC to assure safe facility operation.

a.

Audits of selected aspects of plant operation, as delineated in ANSI N18.7-1976 as modified by the Operational Quality Assurance Plan, shall be performed with a frequency commensurate with their nuclear safety significance and in a manner to assure that an audit of all nuclear safety-related activities is completed within a period of two years. The audits shall be performed in accordance with appropriate written instructions and procedures.

b.

Audits of aspects of plant radioactive effluent treatment and radiological environmental monitoring shall be performed as follows:

1.

Implementation of the Offsite Dose Calculation Manual and quality controls for effluent monitoring at least once every two years.

2.

Implementation of the Process Con:rol Program for solidification of radioactive waste at least once every two years.

3.

The Radiological Environmental Monitoring Program and the results thereof, including quality controls, at least once every year.

c.

Periodic review of the audit program should by performed by the SAC at least twice a year to assure its adequacy.

d.

Written reports of the audits shall be reviewed by the President, NSP Nuclear Generation, by the SAC at a scheduled l

meeting, and by members of Management having responsibility in the areas audited.

6.2 239 Amendment No. 45,46,44,59

7. ' Authority The SAC shall be advisory to the President, NSP Nuclear Generation.

l 8.

Records Minutes shall by prepared and retained for all scheduled meetings of the Safety Audit Committee. The minutes shall be _

distributed within one month of the meeting to the President, NSP Nuclear Generation, the Plant Manager, each member of. l the SAC, and others designated by the Chairman or Vice Chairman. There shall be a formal approval of the minutes.

9.

Procedures t

A written charter for the SAC shall be prepared that contains:

a.

Subjects within the purview of the group.

'i b.

Responsibility and authority of the group.

c.

Mechanisms for convening meetings.

d.

Provisions of use of specialists or subgroups.

e.

Authority to obtain access to the nuclear power plant operating record files and operating personnel when assigned audit functions.

f.

Requirements for distribution of reports and minutes prepared by the group to others in the NSP Organization.

i 6.2 240 Amendment No. 3,46,46 f

~.

B.

Ooerations Committee (OC) 1.

Membership f

The Operations Committee shall consist of at least six (6) regular members drawn from the key supervisors of the onsite supervisory staff. The Plant Manager shall serve as Chairman of the OC and shall appoint a regular member to act as

' Vice Chairman in his absence. Altemates to the regular members shall be designated in writing by the Chairman, or Vice Chairman in the Chairman's absence, to serve on a temporary basis. No more than two alternates shall participate as voting members of the Operations Committee at any one time.

2.

Meeting Frequency

[

The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.

3.

Quorum A quorum shall include a majority of the membership, including the Chairman or Vice Chairman.

4.

Responsibilities - The following subjects shall by reviewed by the Operations Committee:

a.

Proposed tests and experiments and their results.

b.

Modifications to plant systems or equipment as described in the Updated Safety Analysis Report and having nuclear safety significance or which involve an unreviewed safety question as defined in 10 CFR 50.59.

c.

Proposals which would effect permanent changes to normal and emergency operating procedures and any other proposed changes or procedures that are determined by the Plant Manager to affect nuclear safety.

j d.

Proposed changes to the Technical Specifications or operating license.

1 e.

All reported or suspected violations of Technical Specifications, operating license requirements, administrative procedures, or operating procedures. Results of investigations, including evaluation and recommendations to prevent recurrence, will be reported, in writing, to the President, NSP Nuclear Generation and to the Chairman of l

the Safety Audit Committee.

l 6.2 241 Amendment No. 3,47,69 i

f.

Investigation of all Reportable Events and Events requiring Special Reports to the Commission.

g.

Drills on emergency procedures (including plant evacuation) and adequacy of communication with off-site support groups.

h.

All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan and the Security Plan (except as exempted in Section 6.5.F), shall be reviewed with a frequency commensurate with their safety significance but at an interval of not more than two years.

i.

Perform special reviews and investigations, as requested by the Safety Audit Committee.

j.

Review of investigative reports of unplanned releases of radioactive material to the environs.

k.

All changes to the Process Control Program (PCP) and the Offsite Dose Calculation Manual (ODCM).

5.

Authority The OC Shall be advisory to the Plant Manager. In the event of disagreement between the recommendations of the OC and the Plant Manager, the course determined t'y the Plant Manager to be the more conservative will be followed. A written summary of the disagreement will be sent to the President, NSP Nuclear Generation and the Chairman of the l

SAC for review.

6.

Records Minutes shall be recorded for all meetings of the OC and shallidentify all documentary material reviewed. The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the President, NSP Nuclear Generation and others designated by OC Chairman or Vice Chairman.

l 7.

Procedures A written charter for the OC shall be prepared that contains; a.

Responsibility and authority of the group.

b.

Content and method of submission of preseniations to the Operations Committee.

6.2 242 Amendment No. 45,25,46

c.

Mechanism for scheduling meetings d.

Meeting agenda e.

Use of subcommittee f.

Review and approval, by members, of OC actions g.

Distribution of minutes

[

S.3 SoecialInsoections and Audits A.

An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified offsite Northern States Power Company personnel or an outside fire protection consultant.

B.

An inspection and audit by an outside qualified fire protection consultant shall be performed at intervals no greater than three years.

6.4 Action to be Taken if a Safety Limit is Exceeded 1

I i

If a Safety Limit is exceeded, the reactor shall be shut down immediately. An immediate report shall be made to the Commission and to the President, NSP Nuclear Generation or his designated alternate in his absence. A complete analysis of the circumstances _

l l

I leading up to and resulting from the situation, together with recommendations by the Operations Committee, shall also by prepared.

This report shall by submitted to the Commission, to the President, NSP Nuclear Generation and the Chairman of the Safety Audit l

Committee within 14 days of the occurrence.

i Reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.

i i

6.2 - 6.4 243 Amendment No. 3

B.

Radiological 1.a. A Radiation Protection Program, consistent with the requirements of 10 CFR 20, shall be developed and followed. The Radiation Protection Program shall consist of the following:

(1) A Radiation Protection Plan, which shall be a complete definition of radiation protection policy and program (2) Procedures which implement the requirements of the Radiation Protection Plan The Radiation Protection Plan and implementing procedures, with the exception of those non-safety related procedures governing work activities exclusively applicable to or performed by health physics personnel, shall be reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager. Health physics procedures not reviewed by the Operations Committee shall be reviewed and approved by the General Superintendent Radiation Serivices.

b.

In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.1 Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

(1) A radiation monitoring device that continuously indicates the radiation dose rate in the area.

(2) 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 rates in the area have been determined and personnel have been made knowledgeable of them.

(3) An individual qualified in radiation protection procedures with a radiation dose rate monitoring device. This individualis responsible for providing positive radiation protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable Radiation Work Permit.

c.

The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition doors sha!! be locked or attended, to prevent unauthorized entry into these areas and the keys or key devices for locked doors shall be maintained under the administrative control of the Plant Manager.

1.

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the Radiation Work Permit issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas. This footnote applies only to high radiation areas of 1000 mrem /hr or less.

6.5 244a Amendment No. 40,39,78

E.

Offsite Dose Calculation Manual (ODCM)

The ODCM shall be approved by the Commission prior to initial implementation. Changes to the ODCM shall satisfy the following requirements:

1.

Shall by submitted to the Commission with the Semi-Annual Radioactive Effluent release report for the period in which the change (s) were made effective. This submittal shall contain:

a.

sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplementalinformation. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a revision date, together with appropriate analyses or evaluations justifying the change (s).

b.

a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c.

documentation of the fact that the change has been reviewed and found acceptable by the Operations Committee.

2.

Shall become effective upon review and acceptance by the Operations Committee.

F.

Security Procedures shall be developed to implement the requirements of the Security Plan and the Security Contingency Plan. These implementing procedures, with the exception of those non-safety related procedures governing work activities exclusively applicable to or performed by security personnel, shall be reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager. Security procedures not reviewed by the Operations Committee shall be reviewed and approved by the Superintendent, Security.

l G. Temocrarv Chanaes to Procedures Temporary changes to those procedures which are required to be reviewed by the Operations Committee described in A, B, C, D, E and F above, which do not change the intent of the original procedures may be made with the concurrence of two members of the unit management staff, at least one of whom holds a Senior Operator License. Such changes should be documented, reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager within one month. Temporary changes to health physics and security procedures not reviewed by the Operations Committee shall be reviewed by the General Superintendent, Radiation Services for health physics procedures and the Superintendent, Security for security procedures.

6.5 246b Amendment No. 45,25,39,68

j-

^

4 L

r h

B. - Becords Retained for Plant Life (continued)

i

11. Records of the service lives of all safety-related snubbers, including the date at which the service life commences and i

associated installation and maintenance records.'

i L

... p i

-[

r i

I i.

i 1

?

.?

t i

t t

i s

4 F

6.6

.247a 1

Amendment No.~ 9

+

t

.+

1 k

B.

Reoortable Events The following actions shall be taken for Reportable Events:

i l-a.

The Commission shall be notified by a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50 -

and,

- b.

Each Reportable Event shall be reviewed by the Operations Committee and the results of this review shall be submitted to the Safety Audit Committee and the President, NSP Nuclear Generation.

l

~;

i I

i e

a

?

7 3

L

+

i i

i 1

6.7 250

' Amendment No.45,46 i

i t

-~

=

.

ML20154L955

Text

Navigation menu

Search