ML20024G269
| ML20024G269 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/30/1977 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20024G266 | List: |
| References | |
| NUDOCS 9102080454 | |
| Download: ML20024G269 (44) | |
Text
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EX11IDIT B LICEICE AMENIEENT RIQUEST DATED AUGUST 30, 1977 Exhibit B, attached, consists of revised and newly prepa nd pages of the Appendix A Technical Specifications as listed below. Ibese pages incorporate the proposed changec.
Pages 11 iii g
V v11 ix 88 89 90 94 96 97 98 99 100 101 102 103 104 105 106 107 108A 114 119 120 122 -.29 (rents:ioered pages 121A - 123) 135 136 137 138 146 151 l
152 189R (new page) 189S (new page) 189r (new page) l 9102000454 770030 l
PDR ADOCK 05000263 p
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49 E.
Beactor Building Ventilation Isolatior armi Standby Cas Treatment System Initiation 64 32 Bases i
71 4.2 Bases 75 3 3 arx14 3 Control Bod System 75 A.
Reactivity Limitations 76 B.
Control Ibd Withdrawal 79 C.
Scram Insertion Times 80 D.
Control Bod Accumulators 81 E.
Reactivity Anomalies 3 3 and h.3 Ibses 82 88 3 4 and 4.4 Standby Liquid Control System 88 A.
Normal Operstion 89 B.
Operation with Inoperable Components 90 C.
Volume-Concentration Bequirements 94 3 4 and 4.h Bases 96 3 5 and 4.5 Core and Containment Cooling Systems 96 A.
Core Sprsy System 99 B.
LPCI Subsystem 101 C.
FJIR Service Water System 11 REY W
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T 103 D.
IIPCI System E.
Automatic Pressure Relief System 105 106 F.
RCIC System 107 Minimum Core and Containment Cooling System Availability G.
11.
Deleted 108A 1.
Recirculation System 109 3.5 Bases 4.5 Bas es i
115 3.6 and 4.6 Primary System Boundary I
115 i
A.
Reactor Coolant lientup and Cooldown l
115 I
Peactor Vessel Te=perature and Pressure B.
ll6A 1
C.
Coolant Chemistry l
118A l
D.
Coolant Leakage l
119 E.
Safety / Relief Valves F.
de le ted 120 i
G.
Jet Pumps l
121 11.
Ilydraulic Snubbers 130 3.6 and 4.6 Bases til-FEl
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Standby Diesel Generators 182 4.
Station Battery Systems 183 3.9 Bases 185 4.9 Bases 186 3.10 and 4.10 Refueling 187 A.
Refucting Interlocks 187 B.
Core Monitoring 188 C.
Fuel Storage Pool Water level 188 D.
Movement of Fuel 188 E.
Extended Core and Control Rod Drive Maintenance 188A' i
3.10 and 4.10 Bases 189 1
3.11 and 4.11 Reactor Fuel Assenblies 1898 A.
Average Planar Linear lleat Generation Rate 189B B.
Local IJIGR 189C C.
Minimum Critical Power Ratio 189D t
3.11 Bases 189E 4.11 Bases 189G 3.12 and 4 12 Scaled Source Contamination 189N A.
Contamination Ir9N B.
Records 189P 3.12 and 4.12 Bases 189Q 313 and 4.13 Inservice Inspection rex 1 Testira; 189a 313 and h.13 Enses 1897 Y
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LIST OF FIGURES Figure No.
Page No.
2.1-1 Deleted 2.3.1 APRM Flow Referenced Scram and Rod Block Trip Settings 2.3.2 Relationship Between Peak liest Flux and Power for Peaking Factor of 3.08 12 4.1.1
'M' Factor - Graphical Aid in the Selection of an Adequate Interval Between Tests 46 4.2.1 System Unavailability 74 3.4.1 Sodium Pentaborate Solution Volume - Concentration Requirements 92 3.4.2 Sodium Pentaborate Solution Temperature Requirements 93 3.6.1 Change in Charpy V Transition Temperature versus Neut ron Exposure 123 126 3.6.2 Minimum Temperature versus Pressure for Pressure Tests 3.6.3 Minimus Temperature versus Pressure for Mechanical lleatup or Cooldown Fo110 wing W
Nuclear Shutdown 3.6.4 Minimum Temperature versus Pressure for Core Operation 126 l
h 4.6.2 Chloride Stress Corrosion Test Results 8 500*F 129 a
3.II.1-A Maximum Average Linear lleat Generation Rate versus Planar Average Exposure Monticello 8D219 Fuel 18911 3.11.1-B Maximus Average Linear lleat Generation Rate versus Planar Average Esposure Monticello 7D230 Fuel 1891 vil REV E
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I LIST OF TABLES 3.1.1 Reactor Protection System (Scram) Instrument Requireseats 30 4.1.1 Scram Instrument Functional Tests - Minimum Functional Test Frequencies for Safety Instrumentation and Control Circuits 34 4.1.2 Scram Instrument Calibration - Minimum Calibration Frequencies for Reactor Protection Instrument Otannels 36 3.2.1 Instrumentation that Initiates Primary Containr -^
Aion Functions 50 3.2.2 Instrumentation that Initiates Emergency Core Cooling Systems 53 3.2.3 Instrumentation that Initiates Rod Block 57 4
3.2.4 Instrumentation that Initiates Reactcr Building Ventilation Isolation and Standby Gas Treatment System Initiation 60 3.2.S Trip Functions and Deviations 69 4.2.1 Minimum Test and Calibration Frequency For Cora Cooling Rod Block and Isolation Instrumentation 61 3.6.1 Safety Related Ily3raulic Snubbers 123 i
}
3.7.1 Primary Cor :-
Isolation 153 r
4.8.1 Sample Colle,
.c
_sd Analysis Monticello Nuclear Plant - Environmental 1
Monitoring Program 174 6.5.1 Protection Factors for Respirators 206 s
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s 3.0 LIMITING CONDITIONS FOR OPERtJION h.O SURVEILLANCE REQUIPD!alTS l
3.4 STANDBY LIQUID CONTROL SYSTD1 L.h STANDBY LIQUID CONTROL SYSTD4 Applicability:
Aplicability:,
9 Applies to.the operating status of the Applies to the pe.
die testin6 require-standby liquid control system.
ments for the sta. Ny liquid control system.
l Objective:
1 Objective:
To assure the availability of an To verify the operability of the standby independent reactivity control mechanism.
liquid control system.
SPECIFICATION:
SPECIFICATION:
l A.
Normal Operation A.
The operability of the standby 11guld l
3 control system shall be verified by l
l
- 1. The standby liquid control system perforsance of the following tests:
l shall be operable at all times when fuel is in the reactor and the reactor is not shutdown by control 1.
At least once each opersting cycle rods, except as specified in 3.h.B.
r*n"allv initiate one of the two standby liquid control sys'.ots and
- 2. Each standby liquid control system pu=p pu=p d etneralized water into the shall be capable of delivering 24 gpm reactor vessel. Both systems shall against a reactor pressure of 1275 peig.
be tested and inspected in the course of two operating cycles.
3 The system pressure relier valves shall be operable with a setpoint between 1350 and 1450 peig.
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30 LIMITING CONDITIONS FOR OPERATION k.0 SURVEIIJANCE RBQUIRENDfTS
- 2. Inservice inspection and testing of cot:penents shall be conducted in accordance with Specifiestion 4.13 B.
Surveillance with Inoperabic Cocponents B.
Operation with Inoperable Components From and after the date that a redundant When a cocponent beco:nes inoperable, its redundant co=ponent shall be demonstrated coc:ponent is ende or found to be inoperable, Specification 3.h.A shall be conside nd to be operable iz=nediately and daily thereafter.
fulfilled, provided tint the component is returned to an oparable condition within seven days.
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9 30 LIMITING CONDITIONS FOR OFERATION u.O SURVEIILANCE REQUIRENDffS C.
Volune-Cencentation Requirements C.
'Ibe availability of the proper boron bearing solution shall be verified by The liquid poison tank shall contain performnce of the fol4cving tests:
a boron bearing solution that satisfies the volume-concentration requirements 1.
At least once per month -
h of F16ure 3.h.1 and at all times when the stardby liquid poison systen is r*
Boron concentration shall be quired to be operable the temperature detemined.
In addition, the bomn shall not be less than the solution to:D-concentration shall be detemined ersture presented in Figure 3.4.2.
In any time water or boron are added addition, the heat tracing on the pump or if the solution te=perature drops suction lines shall be operable whenever below the limits specified by the room temperature is less than the Figure 3.4.2.
solution temperature presented in Figure 3.4.2.
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Bases 3.4 and 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 con-trol 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 900 ppm of boron in the reactor core in less than 125 minutes.
900 ppm boron concentration in the reactor core is required to bring the reactor from full power to a 37. A k suberitical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 257. boron concentration margin for possible imperfect mixing of the chesaical solution l
in the reactor water and dilution from the water in the cooldown circuit. A miniatus net quantity of 1400 gallons of solution having a 21.47. sodium pentaborate concentration is required to meet this shut-down requirement.
h 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 menon poison peak. The maximum net storage volume of the boron solution is 26)$ gallons. (256 gallons are contained below the pump suction and, therefore, have not been used in the net quantities above.)
Boron concentration, solution temperature, and volume (inclu.fing check of tank heater and pipe heat tracing system) are checked on a frequency to assure a high reliability or operation of the system should it ever be required. Experience with pump opernbility demonstrates that testing at a three-conth inte. val is adequate to detect if failures have occurred.
Standby liquid contml system components are inep-eted and tested in r.ccordance with the requirements of 10 CFR 50, Section 50 55a(g). These requirements are delineated in Specification h.13 This inspection and testing program, combined with the additional surveillance requirements contained in this section, pmvide a high degree of assurance that the standby liquid contml system wila perfom as required when needed.
The relief valves in the standby liquid control system protect the system piping and positf we dis-placement pumps which are nominally designed for 1500 psi from overpressure. The pressure relief valves discharge back to the standby liquid control solution tank.
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i 30 LIMITING CONDITIONS FOR CPERATION h.O SURVEIIIANCE REQUIRDG5ffs i
i 3.5 CORE AND CONTAINMENT COOLING SYSTDG k.5 CORE AND CONTAINMDIT COOLING SYSTIDES l
Applicability:
Applicability:
Applies to periodic testing of the amergency 9
Applies to the operational status of the emergency cooling syste=s.
cooling systems.
Objective:
Objective:
+
1 1
To insure adequate cooling capability for heat To verify the operability of the emergency i
l removal in the event of a loss of coolant cooling systems.
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r.ccident or isolation from the normal reactor heat sink.
Specification:
Specification:
a Low Pressure Core Cooling Capability Ier Pressure Core Cordime Casability A.
Core Spray System A.
Surveillance of the core sp:37 system shell be performed as follows:
1.
Except as specified in 3 5. A.2.,
3 5.A.3., and 3 5.A.5. below, both core 1.
Routine Testing i
spray subsystems shall be operable when-ever irradiated fuel is in the reactor
- a. A sinulated autoentic actuation test vessel and reactor coolant water tempera-shall be conducted each refueliry outage.
ture is greater than 212 F.
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- b. Core spray headerap instrumentation shall be checked once each day, tested l
once each month, and calibrated once
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each 3-month period.
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30 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMBETS c.
Inservice inspection and testing of components shall be conducted in 4
acconlance with Specification 4.13 I
I 2.
From an:1 after the date that one of the 2.
When it is determined that one core core spray systems is made or found to be spray system is inoperable, the ope >
inoperable for any reason, reactor opern-able core spray system ami the LPCI tion is permissible only during the suc-mode of the RER system ami the diesel ceedirg fifteen days unless such system generators required for operation of is sooner made operable, provided that such components (if no external source during such fifteen days all active compo--
of power were available) aball be nents of the other core spray system and demonstrated to be operable 1 6 -
the LPCI mode of the RHR system ami the tely. The operable core spruy system diesel generators required for operation shall be demonstrated to be operable of such components (if no external source daily thereafter.
of power were available) shall be operable.
3 Fne and after the date that both core 3
When it is determined that both core spray systems are made or found to be spray systems are inoperable, the inoperable for any reason, reactor LPCI mode of the RER system ani the 3.S/h.5 97 Rtv H
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h.O GUBVEILIAUCE EEQUIRC4ENTS 30 LIMITING CONDITIONS FOR OPERATION diesel generators required for l
operation is pemissible only during operation of such components (if the succeeding seven days unless at no exterral source of power were least one of such systems is sooner available) shall be demonstrated made operable, provided that during to be eperable irmiediately and such seven days all active components of the LPCI mode of RiiR cyste= and the daily thereafter.
g diesel genenstors required for operation of such components (if no external source of power were available) shall be opera-ble.
4.
Each core spray system shall be capable of delivering 3,020 cpm against a reactor pressure of 130 psig.
If this rate of delivery requirement cannot be met, the system shall be considered inoperable.
S-If the mquimments of 3 5. A.1 - 3 cannot be met, an 'onlerly shutdown of the reactor vill be initiated and the reactor water temperatum sha'l be reduced to less than 212 0F vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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30 LIMITING CONDITIONS FOR OPERATION k.O SURVEIILANCE REQUIRDUNTS j
5 B. Low Pressure Coolant Injection (LPCI) Subsystem B. Surveillance of the Iov Preusure Coolant i
j (LPCI mode of FER system)
Injection (LPCI) Subsystem (LPCI mde of i
RER system) shall be perfomed as follows:
i 1.
Except as specified in 3 5.B.2 and
vhenever irradiated fuel is in the reactor
- a. A simulated autmatic actuation test
- l
' vessel and reactor coolant temperature is shall be conducted each refuelin6 outage.
k h )
I greater than 212 F.
- b. Inservice inspection and testing of i
components shall be conducted in acconinnce with Specification 4.13 b
- c. During each five year period, an air f
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test shall be perfortsed on the dryvell l
spray headers and nozzles.
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I 2.
From and after the date thnt one of the 2.
When it is determined that one of the
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LPCI pumps or ad=ission valves is mde LPCI pu=ps is inoperable, the remaining i
or found to be inoperable for any reason, active components of the LPCI and con-reactor operation is pemissible enly tain=ent coolin6 subsystem, both core during the succeeding thirty days unless spray syste=s and the diesel generstors l
s such pump 'or admission valve is sooner required for operation of such components rnde operable, provided that during such (if no external source of power were j
thirty days the remaining active co=penents available) shall be demonstrated to be I
of the LPCI and containment cooling sub-operable irmediately and the operable j
system and all active co=ponents of both LPCI pu=p daily thereafter.
l core spray systems and 'he diesel genera-tors required for operation of such cos-I ponents (if no external source of p<.,ver were available) shall be operable.
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30 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEIIIANCE REQUIRB4ENTS 3
From and after the date that two of the 3
When it is detemined that the LPCI LPCI pumps or admission valves are made subsystem is inopemble, both core or found to be inopemble for any reason, spray systems, the containment cooling i
reactor operation is pemissible culy subsystem, and the diesel generators g
during the succeeding seven days unless required for operation of such con-such pumps or admission valves are made ponents (if no external source of operable sooner, provided that during such power were available) shall be seven days all active components of both demonstmted to be operable irmnedi-core sprey systems, the containment stely and daily thereafter.
cooling subsystem (including 2 LPCI pumps) and the diesel generators required for operation of such components (if no external source of power were available) shall be demonstrated to be openble at least once each day.
4 A maximum of one dryvell spray loop (containment cooling :nle of RHR) may be inoperable for 30 days when the reactor water temperature is greater than 212 F.
If the loop is not returned to service within 30 days, the orderly shutdown of the reactor vill be initiated and the reactor water te=pereture shall be 0
reduced to less than 212 F.
5 Each LPCI subsystem (RHR) punp shall be capable of delivering h,000 gpm against a reactor pressure of 20 psig.
If this 100 3 5/4 5 REV
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I 30 LIMITING CONDITIO?iS FOR OPERATION h.O SURVEILIANCE REQUIRCICiTS rate of delivery requirement cannot be met, the pump shall be considered inoper-able.
6.
If the requirements of 3 5.B.14 cannot be met, an orderl,y shutdown of the reactor vill be initiated and the reactor water temperature shall be reduced to less than D
212 F vithin 2k'h5ars.
Containment Cooling Capability Containmant Cooling Capability C.
Residual Heat Removal (RHR) Service 'Jater System C.
Surveillance of the RER service water system shall be performed as follows:
1.
Except as specified in 3 5.C.2 and 3 5.C.3 below, both RHP service vster system loops 1.
Inservice inspection and testing or shall be operable whenever irradiated fuel components shall be conducted in is in the reactor vessel and reactor coolant accordance with Specification 4.13 te=perature is greater than 2120F.
2.
From and after the date that one of the 2.
RHR service water system pumps is ende or "Jhen it is determined that one RHR found to be inoperable for any reason, service water pu=p is inoperable, the redundant comconents of the 101 3 5/h.5
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30 LDGTING CONDITIO!G WR OPERATION k.O SUINFTTJANCE RIBEUIRDENTS reactor operation is pemissible only remaining subsysten shall be during the succeeding thirty days unless demonstrated to be operable inenedi-such pump is sooner made opernble, pro-ately and datlv thereafter.
[
vided that during such thirty days all other active components of the RHR service water system are opersble.
O 3
From and after the date that one of the 3
When one RHR service water system RHR service water systems is made or found becomes inoperable, the operable to be inoperable for any reason, reactor system shall be demonstrated to be operution is pemissible only during the operable imediately and daily succeeding seven days unless such system thereafter.
is sooner made opersble, provided that during such seven days all active compo-nents of the operable RHR eervice water
[
system shall be demonstrated to be opera-ble at least once each day.
4 To be considered operable, a RHR service l
water pump shall be capable of delivering 3500 gpm against a head of 500 feet.
i 5
If the recuirements of 3 5.c.1-3 cannot i
be met, an orderly shutdown of the reactor h;
vill be initiated and the reactor water temperature shall be reduced to less than 2120F within 1% hours.
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3.0 LIMITING CONDITIOUS FCR OPERATION h.0 SURVEILINEE REQUIRE lMS High Pressure Core Cooling capability High Pressure Core Cooling Capability D.
High Pressure Coolant Injection (HPCI) System D.
Surveillance of HPCI Systet shall te performed as follows:
9 1.
Except as specified in 3.5.D.2 below, 1-T1outine Testing the HPCI system shall be operable shen-ever the reactor pressure is greater than
- a. A simulated aut - tic actuation test 150 psig ani irradiated fuel is in the shall be conducted each refueling j
reactor vessel, outage.
i
- b. Inservice inspection and testing of ecn::ponents shall be conducted in accordance with Specification h.13 4
l 2.
From and after the date that the HPCI 2.
When it is determined that HPCI system is made or found to be inoperable system is inoperable, the RCIC system, for any reason, reactor creration is per-the LPCI subsystem, and both of the core missible only durine,the succeedin.: seven spray systems shall be demonstrated days unless such system is sooner made to be operable icnediately.
operable, provided that during such seven days all of the Automatic Pressure Relief system, the PCIC system, both of the core
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spray systems, and the LPCI subsystem and containment cooling rode of the PJiR system are operable.
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4 3.0 LIMITING CONDITIONS FOR OPERATION k.O SURVEILIAUCE REQUIREP.rJITS 3
To be considered operuble, the HPCI system shall meet the following conditions:
'Ibe IIPCI shal.1 be capable of delivering a.
3,000 gpm into the reactor vessel for gW the reactor pressure range of 1120 psig to 150 psig.
b.
The condensate storage tanks shall contain at least 75,000 gallons of condensate water.
c.
The controls for automatic transfer of the HPCI pump suction from the condensate storage tank to the suppression chamber shall be operable.
h.
If the requirements or 3 5.D.1-2 cannot be met, an orderly reactor shutdown shall be initiated insnediately and the reactor pressure shall be reduced to 150 psis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
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30 LIMITING CONDITIONS FOR CPERATIOg 4.0 SURVEILIAI:CE REQUIRDiEITS E.
Automatic Pressure Relief System E.
Surveillance of the Automatic Pressure Relief System shall be performed as 4 -
follows:
1.
Except as specified in 3 5.E.2 and
- 1. Routine Testing 3 5.E.3 below, the entire automatic pressure relief syctem shall be operable
- a. A simulated automatic actuation at any time the reactor pressure is above test shall be conducted each oper-l l 150 psig and irradiated fuel is in the ating cycle.
reactor vessel.-
[ ls
- b. Ece each operating cycle, valve oper-__
l it ability shall be verified by cycling 2.
From and after the date that one of the the valves and observing a compenaating automatic' pressure relief system valves is change in turbine bypass valve _ position.
i tede or found to be inoperable for any reason," reactor operation is permissible Inservice inspection and testing o'f ~
c.
only during the succeeding seven days components shall be conducted in unless such valve is sooner made operable, a.cordance with Specification 4.13.
I provided that during such seven days both remaining autocatic relief system valves and the HPCI system are operable.
3 From and after the date that more than
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one of the automatic pressure relief valves are cade or. found to be inoperable
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for any reason, reactor operation is 2.
When it is determined that one or permissible only during the succeeding more automatic pressure relief velves 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unlecs repairs are made and of the Autonsatic Pressure Relief
.provided that during such time the FPCI system is inoperable, the HPCI systema system is crerable.
shall be desmonstrated to be operable iremediately and weekly thereafter.
4 If the requirements of 3 5.E.1-3 cannot be met, an orderly reactor shutdown shall be initiated ir: mediately and the reactor shall be reduced to 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
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4.0 SURETTJANCE REQUIRDENf8 3.0 LIMITING CONDITIONS FOR OPERATION Reactor Core Isolation Cooling System (RCIC)
F.
Surveillance of Reactor Core Isolation Cooling System (RCIC)
F.
Surveillance of the RCIC System shall be performed as follows:
l
)
- 1. Routine hating 1.
Except as specified in 3.5.F.2 below, the RCIC system shall be operable whenever
- a. A simulated automatic actuation test the reactor pressure is greater than 150 shall be coniucted each refueling outa68.
psig and irradiated fuel is in the reactor
- vessel,
- b. Inservice inepection aal testing To be considered operable, the RCIC of components sihall be conducted in accordance with Specification h.13 a.
system shall be capable of delivering LOO gpm into the rtsctor vessel.
2.
From and after the date ttnt the RCIC sys-2.
'4 hen it is determined that the RCIC sys-
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tem is made or found to be inoperable for tem is inoperable, the HPCI system shall
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I any reason, reactor operation is permissible be demonstrated to be operable im:r.ediately only during the succeedinc.15 days unless and daily thereafter.
such system is sooner cade operable, provided that during such 15 days all active compo-nents of the HPCI system are opemble.
3 If the requiremnts of 3 5.F.1 - 2 cannot be met, an orderly shutdown of the mactor eDS11 De initiated irDediately,and tbc reactor pressure shall be nduced to 150 peig g
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the mefter.
REV 35A5
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.M 30 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE BEQUIR N o G.
Minim e Core and Containment Cooling System G.
Surveillance of Core and Containment Availability Cooling System 1.
During any perial when one of the standby 1.
When it is determined that one of h
diesel generators is inoperable, continued the standby diesel generators is reactor operation is permissible only inoperable, all low pressure core during the succeeding seven days, provided cooling and containment cooling that all of the low pressure core cooling service water systems connected to and conts,inment cooling subsystems connee-the operable diesel generator shall ted to the operable diesel generator shall be demonstrated to be operable in--
be operable. If this requirement cannot mediately and daily thereafter.
In be met, an orderly shutdown shall be addition, the operable diesel gen-initiated and the reactor water temperature erator shall be demonstrated to be shall be reduced to less than 2120F vithin operable immediately and daily 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
thereafter.
2.
Any combination of inoperable components in the core and containment cooling systems shall not defeat the capability of the remaining operable components to fulfill g
the core and containment cooling functions.
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-y 3.0 LIMITING CONDITIONS FOR OPEPATION 4.0 SURVEILIANCE REQUIREMENTS i
l I.
Recirculation System I.
Recirculation System 1.
Except as specified in 3.5.1.2 below, whenever 1.
Once per month, when irristed fuel is in the irradiated fuel is in the resctor, with reactor reactor with reactor coolant temperature greater coolant temperature greater than 212 F and both than 212*F and both reactor recirculation reactor recirculation pumps operating, the pumps operating, the recirculation system cross recirculation system cross tie valve interlocks tie valve interlocks shall be demonstrated to be operable by verifying that the cros tie i
shall be operable, valves cannot be opened using the normal control f
2.
The recirculation system cross tie valve inter-switch.
locks may be inoperable if at least one cross tie valve is maintained fully closed.
2.
When a recirculation system cross tie valve interlock is inoperable, the position of at lease one fully closed cross tie valve shall be recorded daily.
3.
Valves in the equalizer piping between the recirculation loops shall be closed. Reactor operation with one loop shall be limited to 3
Inservice inspection and testing of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
components shall be conducted in accordance with Specification 4.13 I
3.5/4.5 108A REV
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l Bases 4.5:
The testing interval for the core and containment cooling systems is based on a quantitative reliability analysis, judgment, and practicality. The core cooling systems have nut been designed to be fully testable during operation. For example, the core spray final admission valves do not open until reactor pressure has fallen to 450 psig; thus, during operation even if high dryvell pressure were simulated, the final valves would not open. In the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel, which is not desirable.
g The systems can be automatically actuated during a refueling outage and this will be done. To increase the availability of the individual components of the core and containment cooling systems, the components which make up the system, i.e., instrumentation, pumps, valve operators, etc., are tested more frequently.
The instrissentation will initially be functionally tested once per month until a trend is established and thereafter according to Figure 4.1 (see Section 3.1/4.1) with an interval not greater then three months.
Core and containment cooling system components are inspected and tested in accordance with the requirements of 10 CFR 50, Section 50.55a(g). These requirem-nts are delineated in Specification 4.13 This inspection and testing program, combined with the additional surveillnnee requirements contained in this section, provide a high degree of assurance that the core and containment cooling systems will perform as required when needed.
With components or subsystems out-of-service, overall core and contairunent cooling reliability is main-tained by demonstrating the operability of the remaining cooling equipment. The degree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment. For routine out-of-service periods caused by preventative maintenance, etc., the pump and valve opert.bility checks vill be performed to demonstrate operability of the remaining components. However, if a failure, design deficiency, etc., caused the out -of-service period, then the demonstration of operability should be thorou6h enough to assure that a similar problem does not exist on the remaining components., For estemple, if an out-of-service period were caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test in addition to the operability checks.
t 4.5 BASES 114 REV
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L 4.0 SURVEILUL' ICE REQUIREMDITS 3.0 LIMITING CONDITIONS MR OPERATICN E.
Safety / Relief Valves E.
Safety / Relief Valves 1.
The integrity of the safety /re ler 1.
During power operating conditions and whenever valve bellows shall be continuously reactor coolant pressure is greater than 110 psig monitored.
and temperature is greater than 3' 3 F :
- 2. The opernbili' y of th? bellows The safety valve function (self-actuation) of monitoring system shall be demonstrated a.
seven safety / relief valves shall be operable, at least once every t.hree months.
- 3. Inservice inspection and testin6 of b.
" die solenoid activated relief function
( Automatic Pressure Relier) shall be components shall be conducted in l
acconiance with Specification 4.13 operable as required by Specification 3 5.E.
S 119 3.6/4.6 REV
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-3 3 0. LIMITIT;G CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIRDENTS F.
deleted F.
deleted G
-G.
Jet Pumps G.
Jet Pumps Whenever the reactor is in the Startup Whenever there is recirculation flow with the or Run modes, all Jet pumps shall be oper-reactor.in the Startup or Run modes, Jet pump able. If it is determined that a jet pump is operability shall be checked daily by verify-inoperable, the plant shall be placed in a ing that all.the following conditions'do not cold shutdoun condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
occur simuftaneously:
- i
- {
1.
The two recirculation loop flows are g-unbalanced by l'i% or more when the recirculation pumps are operating at the same speed.
2.
The indicated value of core flow rate is 10% or more less than the value de-rived from loop flow measurements.
l 3 6/4.6 13 azy
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3.@ LIMITING Cor3DITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 5.
Snubbers may be added to safety related The required inspection interval chall not be syst ma without prior License Amendment lene,thened more than one step at a time.
to Taale 3.6.1 provided that a revision to Table 3.6.1 is included with the next Snubbers may be categorized in two groups, license amendment request.
" accessible" or " inaccessible" based on their accecsibility for inspection during reactor operation.
'Ihese two groups any be inspected independently according to the above schedule.
2.
All bydraulic snubbers whose seal materials are other than ethylene prooylene or other material that has been demonstraced to be compatible with the onaret i nr. environment shall b? visually in-spected for, opernbility every 31 days.
l 3.
Once each refueling cycle, a representative sample of 10 hydraulic snubbers or approximately 10% of the hydraulic snubbers, whichever is less, shall be functionally tested for operability including verification of proper piston movement. lock up, and bleed.
For each unit and subsequent unit found inoperable, an additional 10% or ten hydraulic snubbers shall he so tested until no more failures are found or all units have been tested.
Snubbers designated in Table 3.6.1 as being especially difficult to remove or located in High Radiation Areas during shutdown are exempt from this requirement.
4.
Snubbers may be reclassified as being in or out of fligh Radiation Areas during shutdown in Table 3.6.1 based on the most recent radiation survey provided that a revision to Table 3.6.1 is included with the next license amendment request.
3.6/4.6 122 REV
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.n. S- ?xt. 4k E b FIGURE 4. 6. 2 CHLORIDE STRESS CORROSION TEST RESULTS e 500 F 1800 \\ e 10 g FAILURE i5 E E N g Limit EshWeesd hr Cl at Operahng 0, Caucarjahen = SJ -e.4 pre 0.1 NO FAILURE 0$i 1.o 10
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CHLORIDE, pra a J. k % i. Ed. 3.6/4.6 129 nw I
. _ __ q 3-c..-- ___x- _.. -. z e s Bases Continued 3 6 and h.6: 2e safety / relief valves have two functions; i.e. power relief or self-actuated by high pressure. %e solenoid actuated function (Automatic Pressure Relief) in which external instrumentation signals of coincident high dryvell pressure and low-lov veter level initiate opening of the valves. This function is discussed in Specification 3.5.E. In addition, the valves can be operated manually. %e safety function is perfomed by the same safety / relief valve with self-actuated integral bellows and pilot valve causing main valve operation. Article 9 of the AS4E Pressure Vessel Code Section III Nuclear Vessels requires that these bellows be monitored for failure since this vould defeat the safety function of the safety / relief velve. 4 It is realized that there is no vey to repair or replace the bellows during operation and the plant must be shut down to do this. The thirty-day period to do this allows the operator flexibility to choose his time for shutdown; meanwhile, because. of the redundancy present in the design and the continuing monitoring of the integrity of the other valves, the overpressure pressure protection has not been catrprcnised. %e auto-relief function vould not be impaired by a failure of the bellows. However, the self-actuated overpressure safety function vould be impaired by such a failure. Provision also has been made to detect failure of the bellows monitoring system. Testing of this system quarterly provides assurance of bellows integrity. When the setpoint is being bench checked, it is prudent to disassemble one of the safety / relief valves to examine for crud buildup, bending of certain actuator members or other signs of possible deterioration. The program of safety / relief valve testing confoms to the requirements of 10 CFR 50, Section 50 55a(g). 'Ibese requimments are delineated in Specification h.13 mis inspection and testing progmm, combined with the additional surveillance requirements contained in this section, provide a high degree of assurance g that the safety / relief valves vill perform as required when needed. 3.6/h.6 BASES 135 REV
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.- -== smr,my-n -. - yy i, m r c-a v 8 s Bases Continued 3.6 ani 4.6: G. Jet pumps Failure of a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended 31ne break. %erefore, if a failure occurred, repairs must be made. l B e detection technique is as follows. With the tvo recirculation pumps balanced in speed to vithin + 5%, the flov rates in both recirculation loops vill be verified by Control Rocm monitoring instruments. l l Tf the two flow rate values do not differ by more than 10%, riser and nozzle assembly integrity has been verified. If they do differ by 10% or more, the core flow rete measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived frem the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10% or more (vith the derived value higher) diffuser measurements vill be taken to define the location within the vessel of failed jet m nozzle (or riser) and the plant shut dcun for repairs. If the potential blevdown flov area is increased, the system resistance to the recirculation pump is also reduced; hence, l the affected drive pump vill "run out" to a substantially higher flow rate (approximately 115% to 120% for l a sirgle nozzle failure). If the two loops are balanced in flov at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates vould be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past Be reverse flow through the inactive jet pump would still be the core thus reducing the core flow rate. indicated by e positive differential pressure but the net effect would be a slight decrease (3% to 6%) in the total core flow measured. Bis decrease, toEether with the loop flow increase, vould result in a lack of correlation between measured and derived core flow rate. Finally, the affected jet pump diffuser differential pressure signal vould be reduced because the backflow would be less than the nomal forward flow. A nozzle-riser syste:s failure could also generate the coincident failure of a jet ptmp body; however, the converse is not true. We lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle-riser system failure. I l 137 RW 3 6/4.6 BASES
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-oq [ht; L. ' 4 5 Bases Continued 3.6 and 4.6: Shock Suppressors (Snubbers) 11, Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earth-the consequence quake or severe transient, while allowing normal thermal motion during startup and shutdown, of an inoperable snubber is an increase in the probability of structural damage to piping as a reasit of a seismic or other event initiating dynamic loads. It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be operable during reactor operation. low probability events, a period of 72 hours Because the snubber protection is required only duringIn case a shutdown is required, the allowance of 36 hours to reach a cold is allowed for repairs or replacements. Since plant ( shutdown condition will pemit an orderly shutdown consistent with standard operating procedures. Specification 3.6.11.4 prohibits i I startup should not commence with knowingly defective safety related equipment, startup with inoperable snubbers, The in-All safety related hydraulic snubbers are visually inspected for overall integrity and operability. f spection will include verification of proper orientation, adequate hydraulic fluid level and proper attachment of snubber to piping and structures, level of snubber protection. 'Ihus the required The inspection frequency is based upon maintaining a constant The number of inoperable snubbers j inspection interval varies inversely with the observed snubber failures. found during a required inspection determines the time interval for the next required inspection. Inspections intet val has elapsed may be used as a new reference point to determine the next in-performed before that However, the results of such early inspections performed before the original required time interval g spection. has elapsed (nominal time less 257.) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule. Experience at operating facilities has shown that the required surveillar:ce program should assure an acceptable level of snubber perfomance provided that the seal materials are compatible with the operating environment. lab tests or Snubbers containing seat material which has not been demonstrated by operating experience, analysis to be compatible with the operating environment should be inspected more frequently (every soonth) is completed. until material compatibility is confirmtd or an appropriate changcout 138 3.6/4.6 BASES RW
~_ _m- _m_-- -u gg3%QgEW N 'G ^3 {y3-m33 1 : uu ?Q ._ s s-3.0 LIMITING CONDITIONS FOR OPHIATION 4.0 SURVEILLANCE REQUIREMENTS 1 3 Pressure Suppression Chamber - 3 Pressure Suppression Chamber - Reactor Building Vacuum Breakers . Reactor Building Vacuum Bieakers a. Except as specified in 3 7.A.3.b a. The' pressure suppression chamber-reactor ~ below, two' pressure suppression building vacuum breahers and associated in-chamber-reactor building vacuum strumentation including set point shall be breakers shall be. operable at all checked for proper operation every three times when the primary containment months.: integrity is required. The set point of the differential pressure b. Inservice inspection and testing of. instrumentation which actuates the components shall be conducted in pressure suppression chamber-reactor accordance with S ecification 4.13. t building vacuum breakers shall be 0.5 psi. b. From and after the date that ole of the pressure suppression chamber-reactor building vacuum. breaker s is~ made or found to be inoperable for any reason, reactor operaticn is permissible only during the suceed-ing.seven days unless such. vacuum breaker is sooner made operable, provided that the repair procedure does not violate primary containment integrity. 4 146 REV-3 7/h.7 J .m
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9 . M. =G,. 6 3.0 LIMITING CONDITIOIG FOR OPERATION h.O SURVEILIANCE PMUIRDfEfffS d. The fuel cask or irradiated fuel is not being moved within the reactor L building. D. Primary Containment Isolation Valves D. Primary Containment Isolation Valves 1. During reactor power operating conditions, 1. The primary containment isolation valves all isolation valves listed in Table 3.7.1 surveillance shall be perfoceed as tollows: and all primary system instrument line flow check valves shall be operable except a. At least once per operating cycle the as specified in 3.7.D.2. operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and closure times. b. Inservice inspection and testing of cociponente shall be conducted in accordance with Specification 4.13. 3.7/h.7 151 aEv ..... ~ - - _.....-
,w [ A 4 3.0 LIMITDU CONDITIO!G FOR OPERATION h.0 SURVEILLM7CE RbUIRDEfrS 1 I f 1 l l 2., In the event any isolation valv e specified 2. Whenever an isolation valve listed in in Table 3.7.1 becomes inoperable, reactor Table 3.7.1 is inoperable, the position of I operation in the run code may continue at least one fully closed valve in each line provided at least one valve in each line having an inoperable valve shall be recorded having an inoperable valve is closed. daily. 3. If Specification 3. 7.D 1 and 3.7.D.2 cannot g be me t, initiate normal orderly shutdown and have reactor in the cold shutdown i condition within 24 hours. l 1 l l l 3.7/4.7 19 PEV
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4 3:0 LIMITIIU CONDITIONS FOR OPERATION h.O SURVEILIANCE RMUIREMEIES 3 13 INSERVICE INSPECTION AND TESTIIU 4.13 INSERVICE INSPECTION AND TESTIIU Applicabi.11ty: Applicability: Applies to camponents which am part of Applies to the periodic inspection and the reactor coolant pressure boundary and testing of components which are part of their supports and other safety-related the reactor coolant pressure boundary pressure vessels, piping, pumps, and and their supports and other safety-valves. related pressure vessels, piping, pumps, I l and valves. Objective: Objective: To assum the integrity of the reactor To verify the integrity of the reactor coolant pressure boundary and the coolant pressum boundary and the operational readiness of safety-related operational readiness of safety-pressure vessels, piping, pumps, and related pressure vessels, piping, pumps, valves. and valves. Specification: Specification: A. Inservice Inspection A. Inservice Inspection 1. To be considered operable, Quality
- 1. Inservice inspection of Quality Group A, B, and C compor.ents shall Group A, B, and C components.shall be perfonned in accordance with h
satisfy the requirements contained the requirements for ASME Code Class in Section XI of the ASME Boiler and Pressure Vessel Code and appli-1, 2, and 3 components, respectively, l contained in Section XI of the ASME cable Addenda for continued service Boiler and Pressure Vessel Code and of ASME Code Class 1, 2, and 3 compo-applicable Addenda as requimd by - t nents, respectively, except vbere i 10 CFR 50, Section 50 55a(g), except relief has been req;ested from the Commission pursuant to 10 CFR 50, whe m relief has been requested from Section 50 55a(g)(6)(i). the (bemission pursuant to 10 CFR 50, Section 50 55a(g)(6)(1), 3.13/h.13 189R REV
...2. _.__-m ^ o g O s ~. i 30 LIMITUU CONDITIOfB FOR OPEIMTION h.O SURVEILIANCE REQUIPDGNIS l ( ) B. Inservice Testing of Pumps and Valves B. Inservice Testing of Pumps aal Valves 1. To be considered operable, Quality
- 1. Ineervice testing of Quality Group Group A, B, and C pumps and valves A, B, and C pumps and valves shall be shall satisfy the requirements con-performed in acconlance with the tained in Section XI of the ASME Boiler requirements for ASME Code Class 1,
} and Pressure Vessel Code and appli-2, and 3 pumps and valves, respect ively, I cable Addenda for operability of contained in Section XI of the ASME ASME Code Class 1, 2, and 3 pumps Boiler ani Preesure Vessel Code and and valves, respectively, except applicable Addenda as n quired by where relief has been requested from 10 CFR 50, Section 50 55a(g), except the Commission pursuant to 10 CFTt 50, whe m relief has been requested from Section 50 55a(g)(6)(1). the Commission pursuant to 10 CFR 50, Section 50 55a(g)(6)(i). f l l 3 13/h.13 Ws REV ~
u ~ ~. a :_-. _,y pcump " 5.+ 7 e Bases 3.13 and h.13: The inservice inspection and testing program confome to the requirements of 10 CFR 50, Section 50 55a(g). Where practical, the inspection and testing of components classified into NRC Quality Oroups A, B, and C i vill conform to the requirements for ASE Code Class 1, 2, and 3 components contained in Section XI of the ASME Boiler and Pressure Vessel Code. Using Regulatory Guide 1.26, Revision 3, " Quality Group classifications and Standards for Water, steam, and Padioactive-Waste-Containing Components of Nuclear Power Plants," as a guide, all lenticello components have been classified into Quality Gmups. This classification serves as the basis for detemining which ASME Code Class inspection and testing requirements are applicable to a given component. 10 CFR 50, Section 50 55a(g) requires components which are part of the reactor g coolant pressure boundary and their supports to meet the inservice inspection and testing requirements applicable to components classified as ASME Code Class 1. Other safety-related components must meet the inservice inspection and testing requirements applicable to components classified as ASME Code Class 2 or 3 'Ibe inservice inspection pmgmm must be updated at hO month intervals. The pmgram for testing pumps and valves for operational readiness must be updated every 20 months. A description of the updated programs should be submitted to the N3C for review at least 90 days before the start of each period. A suggested fomnt for this description is contained in Appendix A to reference (1). The inservice inspection and testing pmgram must, to the extent practical, comply wit,h the requirements in editions and addenda to the ASME Code that are "in effect" no more than six nonths before the start of the period covered by the updated program. The tem "in effect" means both having been published by the ASME, and having been referenced in paragraph (b) of 3 CFR 50, Section 50 55a. If a code required inspection or test is impractical, requests for eviations are submitted to the Commission in accordance with 10 CFR 50, Section 50 55a(g)(6)(1). _ne infomation specified in Appendix B to reference (1) should be submitted for each deviation j. requested. Deviation requests should, if possible, be submitted to the NRC for review at least 90 days before the start of each period. Deviations identified during an inspection period may be grouped and requested at'the end L each calendar quarter. It is expected that a small number of deviations vill be identified during the inspection period, particularily the first period when new inspection and testing techniques vill be utilized. A requested devietion request may be considered acceptable to the Commission until a fomal disapproval has been received.
References:
- 1. Ictter from D. L. Ziemann, Chief, Operating Reactors Branch No. 2, USNRC, to L. O. Mayer, NSP, dated November 24, 1976.
313A.13 BASES 189T REV
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Tk/.VictorStello NSP g -l Minneapolis, Minnesota 55401 09/01/77 [FLETTEn TNOTORIZ E D P R OP INPUT PORM NUMBER OP COPf ts RECEIVED 26M tOIN AL [UNC LASSIPIE D 0CoPv 3 <g agg oesCalPTION ENCLOSU RE License No. DPR-22 Appl for Amend: Tech Specs proposed change concerning requiring w inservice inspection and testing to be performed i h .-[h, accordance with the examination and testing requirements set forth in Section XI of the ASME 1p Code and Addenda...Notorized 08/30/77 1 1/4" !ACIGOWLEDGED ,'lPLANTNAME: MONTICELLO j,cm 09/02/77 &Ed-SAFETY FOR ACTION /INFORMATION .i BRANCH CHIEF: (7) D Ad i 5 I t i i m
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