ML20024G478

Proposed Tech Specs Correcting Typos,Clarifying Intent of Tech Specs,Correcting Reactor Coolant Chemistry Surveillance Requirements & Revising Reactor Vessel NDTT Surveillance Program

ML20024G478
Person / Time
Site:	Monticello
Issue date:	11/15/1974
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G471	List: ML20024G470 ML20024G474 ML20024G478
References
NUDOCS 9102120532
Download: ML20024G478 (34)

Text

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u.dii- ; b Licenre Amendment hequect Dated Ibvember 15,197h Exhibit B, attached, consists of newly prepared pt.ges for the Appendix A Technical Sporifications as listed below. Thase paEec incorporate the propoced changec.

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, TABLE OF CONTE'.TS Page 1

1.0 DEFINITIONS 6

2.0 S AFETY lit!ITS N~0 LIMITING SAFETY SYSTEM SETTINGS 6

4 2.1 a d 2.3 Feel Cladding Integrity 13 I 2.1 Bases 18 1

2.3 Bases J

2.2 and 2.4 Reactor Coolant Systen 23 2.2 Bases 24 2.4 Bases 26 LIMITING CONDITIONS FOR OPERATION AND 4.0 SURVEILIR;CE REQUIRE!!ENTS 28 f 3.0 Reactor Protection System 28 3.1 and 4.1 31 Bases 37 4.1 Bases 43 L

t Protective Instrumentation 47 3.2 and 4.2

A. Primary Containment Isolation Functions 47 l

B. Emergency Core Cooling Subsystems Actuation 48 C. Control Rod Block Actuatina 48 i

D. Air Ejector Off-Ga* System 48 I

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1 E. Reactor Building; Ventilation Isolatiot; nd h9 Standby Cac Treatment System Initintion 32 Bases 64 L.2 Bares 71 3 3 and h. 3 Cont ml Bod System 75 A. Benetivity Limitations 75 B. Control Rod 'lithdrnwal 76 C. Scram Insertion Times 79 i

2 D. Control Rod Accumulators 80 E. Beactivity Anomalies 81 3 3 and 4.3 Ibses 82 3.h and h.h Standby Liquid Cont rol Systen 88 A. Normal Operation 88 B. Operation with Inoperable Components 89 C. Volume-Ce? centration Requirements 90 3.4 and 4.4 Bases 94 3 5 and 4.5 Core and Containment Cooling Systems 96 A. Core Sprny System 96 i B. LPCI Subsystem 98 i C. FRR Service IInter System 101 {

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D. HPCI System 103 E. Automatic Pressure Relief System 104 F. RCIC System 106 G. Mininum Core and Containment Cooling System Availability 107 H. Deleted I. Recirculation System 108A J. AveraEs Planar 111GR 10BA K. Incal IIIGR 108B 3.5 Bases 109 4.5 Banes 114 3.6 and 4.6 Primary System Bounddry 115 A. Thennal Limitations 115 B. Pressurization Temperature 116 C. Coolant Chemistry 116 D. Coolant Istage 118 E. Safety / Relief Valves 119 F. Structural Integrity 120 G. Jet Pumps 120 3.6 and 4.6 Bases 130 111 PEI

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3.7 and 4.7 Cortain ent Systems 139 t

i A. Primary C ntainment 139

3. Standby Gas Treatment Systen 148 C. Secondary Containment 150 l D. Primary Containment Isolation Valves 151 3.7 Bases 156 4.7 Bases 161 3.8 and 4.8 Radioactive Effluents 168 A. Airborne Effluents it,g i B. Mechanical Condenser Vacuum lhimp 170B C. Liquid Effluents 171 D. Radioactive Liquid Storage 173 E. Augmented Off-Gas System 173

+ L

! F. Environmental Monitoring Program 173A 1 '

i 3.9 and 4.8 Bases 177 j 3.9 and 4.9 Auxiliary Electrical Systems 180 A. Operational Requirements for StartuP 180 '

B. Operational Requirements for Continued Operation 131 i

1. Off site Power (Line) 1g1

2. Offsite Power (Transformers.~ 182 ty bel

3. Standby Diesel Generators 182

4. Station Battery Systems 183 3.9 Bases 185 4.9 Bases 186 3.10 and 4.10 Refueling 187 l

A. Refueling Interlocks 187 B. Core Monitoring 188 i C. Fuel Storage Pool Water Izvel 188 D. Movement of Fuel 188 E. Extended Core and Control Rod Drive Paintenance 188A l 3.10 and 4.10 Bases 189 .

5.0 DESIGN FEATURES 190 6.0 ADMINISTRATIVE COh*rROLS 192  :

6.1 Organization 192 6.2 Review and Audit 195 6.3 Actions to be taken in the Event of an Abnomal Occurrence 201 in Plant Operation '

. 6.4 Action to be taken if a Safety Limit is Exceeded 201 6.5 Plant Operating Procedures 202

6.6 Plant Operating Records 209 6.7 Plant Repo. ting Requirements 211

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I LIST OF TABLES 3.1.1 Resctor' Protection System (Scram) Inst rument Requirrment s 30

! 4.1.1 Scram Inst rument Furictional Tests - Minimum Functional T 'st Frequencies 34 t

, for Safety Instrumentation and Control Circuits '

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4.1.2 Scram Instrument Calibration - Minimum Calibrntion frequencies 'fer 36

Reactor Protection Instrument Channels

3.2.1 Inst rument ation that Initiates Primary Containment isoletion Functions 50 J

3.2.2 Instrument ation that Initiates Emergency Core Cooling 53 Systems 3.2.3 Instrumentation that Initiates Rod Block 57 3.2.4 Instrumentation that Initiates Reactor Building Ventilation 60 Isolation and Standby Gas Treatment System Initiation

, 3.2.5 Trip Functions and Deviations 69 4.2.1 Minimum Test and Calibration Frequancy for Core Cooling, Rod Block 61 and Isolation Instrumentation i

4.6.1 In-Service Inspection Requirements for Mbnticello 124 i 3.7.1 Primary Containnent Isolation 153 4.8.1 Sample Collection and Analysis Monticello Nuclear Plant - Environmental 174 tbnitoring Program i

6.5.1 Protection Factors for Respirators 206 4

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Peas Con- inued :

31 Three AM4 instntment channels are provided for wh nrotection trip system. AETU4's #1 and #3 operate contacts in one subchannel, and AFPd4's (2 and !3 operate contacts in the other subchannel.

AFrC4's S,#5, and #6 are arranged similarly in the other protection trip system. Each protection trip syrtem has one more APFit than is necessary 'o meet the mini =um number required. This allavs the bypussing cf one AFRM per protection trip syn *,em for maintenance, testing, or calibration. Ad-ditional IRM -hannels have also been provided to allev for bypassing of one such channel in each trip system.

The basen for the scram settings for the IPO4, Apr 1, high reactor pressure, reactor lo.r water level, turbino centrol valve fast closure, and turbine rtop valve closure are discussed in Specifications 2 3 und 2.4.

Instrumentation (pressure switches) in the drywell are provided to detect a loss of coolant accidem and initiate the emergency core'~coling equipment. This ir.strwentation is a backup to the water level instrtmentation which is discussed in Specification 3 2.

The control rod drive scram system is designed sc that all of the water which is discharged from the reactor by the scram can be accommodated in the discharge piping. A part of this piping is an instru-ment volume which accoccodates in excess of 32 gallens or water and is the low point in the piping.

No credit was taken for this volrme in the design of the discharge piping as concerns the amount of water which must be accoradated during a scram. During normal operation the discharge volume is erpty; however, should it fill with water, the water discharged to the piping from the reactor could nct be accccmodated which would result in slow scram times or partial or no control rod insertion.

l To preclude this occurence, level switches have been Provided in the instnzment volume which alarm -_

and scram the reactor when the volume of water in the discharge volume receiver tank reaches 32 gallons.

As indicated above, there is sufficient volume in the piping to accommodate the scram without impair-ment of the scram times or amount of insertion of the control rods. This function shuts the reactor down uhile sufficient volume remains to acco":nodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately.

Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of' 31 BASFr 38 REv

4 Bases Centinued:

3.1 condenser vacuum initiates a closure of the turbine , top valves and turbine bypass valves which eliminates the heat input to the condenser. C) are of the turbine stop and bypass valves causes a prescu re t ransient, neutron flux rise, and an 1,ctsase fu surface heat flux. To prevent the clad safety limit from being exceeded if this oc urr. a re sctor scram occurs on turbine stop valve closure. The turbine stop valve closure rcram function alone is adequate to prevent the clad safety limit from being exceeded in the evr nt of a turbine trip transient without bypass.

Reference FSAR Section 14.5.1.2.2 and supplement al information submitted February 13. 1973.

The condenser low vacuum scram is a back-up to t he stop valve clesure scram and causes a scram before the stop valves are closed and thus the resulting e.ransient is less severe. Scram occurs at 2 3" Hg vacuum, stop valve closure occurs at 27" If c vacuum, and bypass closure at 7" Hg vacuum.

Itigh radiation levels in the main steamline tunnel above that due to the normal nitrogen and orfgen radioactivity is t a indication of leaking fuel . A scram is initiated whenever such ra31ation level exceeds ten times normal full pcwer background. The purpose of this scran is to reduce the source of such radiation to the extent necessary to prevent excessive release of radioactive materials.

The main steamline isolation valve closure scran is set to scram when the isolation valves are 6107 closed from full open. This scram anticipttes the pressure and flux transient, which would occur when the valves close. By scramming at thi.s setting the resultanc transient is insignificant. '

Refer .nce Section 14.5.1.3.1 FSAR and supplemental in formation submitted February 13, 1973.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status. Reference Section 7.7.1 FCAR.

The manual scram function is active in all modes, thus providing for a manual means of rapidly ins ertIng control rods during all modes of reacter operation.

" e IR" systen provides protection against excessive power levels and shert reactor periods in the i

3.1 BASES 39 REV

_ _ . _ ___ __ . _ __ __._____.____.-__.____u---i =

.vl Fasas W. ire _d o 31 startmp and intere diate p wer ranges. Ref. Sectier 7.h.h FSAR. A source ranga moniter (CE*4) syrter is also provided to supply additicnal neutrcn level inf cr ation during start-up but has no ser m fur tions. Fef. Section 7.h.3 FSAR. Thus,.the IFJ4 is required in the "Fefuel" and

" Start t p" modes. In the power ranga the AFF24 systa- provides required protectic- Ref. Section 7.h.5.P FSAF. Thus, the IF14 system is not required in the "Run " code. The AFR^ 's cover only the power range, the IP'4's provide adequate coverage in the start-up and intemediate range, and ther~

fore, the AITut's are not required for the "Peruel" cr "Startup" rrodes.

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%e high rewtor pressure, high drywell pressure, and reactor low water level scrams are required l

for all modes of plant operation unless the reactor is subcritical and depressurized. ney are,

! therefore, required to be operational for all modes of reactor operation except in the " Refuel" node wi th the reactor subcritical and reactor temperature less than 212 F as allowed by Note 3. i ihe scram discharge volume high level is required for all modes of plant operation and is required to be eperational for all modes. However, it is permissible for this trip to be bypassed in the "Refuef" mode. In order to reset the safety system after a scram condition, it is necessary to drain t he scram discharge volume to clear this scra= input condition. (h is condition usually followc eny scram, no matter what the initial cause might have been.) In order to do this, this i particular scram function can be bypassed only in the refuel position. Since all of the control

! rods are completely inserted following a scram it is permissible to bypass this condition because l

! a control rod block prevents withdrawal as long as the switch is in the bypass condition for ,

i this function. '

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To perrit plant operation to generate adequate steam cnd pressure to establish turbine seals and  !

condenser vacuum at relatively low reactor power. the main condenser vacuum trip is bypassed untii 600 psig. This byp1tes also applies to the main stean isolation valves for the same reason. Ref.

Section 7.7.1.2 FSAR.

An autematic bypass of the turbine coatrol valve fast closure scram and turbine stop valve closure

! scram is effective whenever the turbine first st1ge pressure is less than 301 of its rated value.

' This irwres that reactor themal power is less than 457. of its rated value. Closure of these valves from such a low initial power level does not constitute a threat to the integrity of any barrier to l the release of radioactive material.

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A. The mini =um functional testing frequency ase1 in this specification is based on a reliability analysis using the concepts developed 'in reference (1). This concept was specifically adopted i

to the one out of two taken twice logic of the reactor protection system for Monticello. The anslysis shows that the sensors r_re primarily responsible for the reliability of the reactor protection system. This analysis makes use of unsafe failure rate experience at conventional and nuclear power plants in a reliability model for the system. An " unsafe failure" is defined  :'

as one which negates channel operabili,ty, and which, due to its nature, is revealed only when i

tha channel is functionally tested or attemp+.s te respond to a real signal. Failures such as blevn fuses, ruptured beurdon tubes, faulted amplifiers, faulted cables, etc., which result in

" upscale" or "downscale" readings on tie reactor instrumentation are " safe" and will be easily recognized by the operators during operation because they are revealed by an alam or a scram.

The 13 semm sensor channels listed in Table '4.1.1 are divided into three groups (A., E., and C. )

and are defined on Tab]e h.l.l.

4 The sensors that make up group (A) are specifically selected from among the whole family of industrial on-off sensors that have earned an excellent reprtation for reliable operation.

Actual history on this class of sensors operating in nuclear power plants shows four failures l in '+72 sensor years, or a failure rate of 0.97 X 10-6/hr. During design, a goal of 0.99999 prehability of success (at the 507. confidence level) was adopted to assure that a balanced I

and adequate design is achieved. The probability of success is primarily a function of the sensor failure rate and the test interval. A three-raonth test interval was plan aed for group (A) sensors. This is in keeping with good operating practice, and satisfies the design goal for the logic configuration utilized in the neactor Protection System.

4 To satisfy the long-tem objective of maintaining an adequate level of safety throughout the plant i i

lifetime, a minimum goal of 0 9999 at the 957 confidence level is proposed. With the (1 out of 2)

] X (2) logic, this requires that each senser have an availability of 0 993 at the 95% confidence  !

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.1 r1ASES 43 RE7

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h.1 hour monitoring interval for the analog device ~ as assume i above, and a weekly test interval for the bi-stable trip circuits, the decipi rellability grai af 0.o9999 is attained with arple margin. The test frequency of once per week hoc d zeloped principally on the basis of past prac* ice and r >od judgment, anu nn+ hing has developed t - in';cate that the frequency should change.

Group (C) devices are active cnly during a giv n pertion of the operational cycle. For exar ple, the 1511 is active during startup and inactive urin full-power operation. Thus, the only test that is Franingful is the one perforred just prior to shutdown or startup; 1.e., the tests that are r"rformd just prior to use of the instrum,nt.

Call! mti"n frequency of *he instrument channnt ir livided into two groups as defined on Table h.1."

F_xperience with passive type instruments indic~tes that a yearly celibration is adequate. For those devices which employ a=plifie rs,etc. , dri ft crecifications call for drift to be less than 0 5 1/ month; i.e., in the period of a menth a Jrif t of 0 5% vould occur and thus provide for adequate margin. Fcr the APfet system drift of elec' ronic apparatus is not the only consideration in determining a calibration frequency. Change in Twer distribution and loss of chamber sensitivity dictate a calibration every three days. Calibration on this frequency assures plant operation at or below thermal limits.

B. The peak haat flux shall be checked once per day to determine if the APRM scram requires adjust-ment. This vill normally be done by checking i he ITret readings. Only a smil number of control .

rods are moved daily and thus the peaking facters are not expected to change significantly, thus a daily check of the peak heat flux is adeqt tr .

(1) Reliability of Engineered "lafety Features as a Fun 9 ion of Testing Frequency. I. M. Jacobs, Nuclear S ifety, Vol. 9, No. h , July-Aug. l'V30, pgs. 310-312.

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Tnble 3.2.3 Centirued Notes:

(6) Upon diccovery that minimum requirements for the rneber of oparable or oparating trip systems or instr, ment channels are not entisfied actions shall ha initinted to:

(n) Sa t$ sis, the requirements by placing appropriate chaineln or systems in the tripped condition, er (b) Place the plant under the specified required cond1+. ions using nomal oparating proemiures.

(7) 'Iher- mint be n totnl of at least !4 operable er operating AP'34 channels.

• Ibquired ennditioun when minimum conditions for operation nra no*. satisfied.

A. Ibsetor in Shutdown mode.

B. No rod withdrawnls pamitted while in Befual or Star

• up rode.

C. Fre" tor in Ibn mode.

D. No rol withdrawals permitted while in the Ihn mode.

E. Ihr on I!M range oc below atrl renetor in Startup, "crual, or Shutdovri mode.

• • Allovnble Pypass Conditions

n. SIM Detector-not-fully-inserted rod block any be bypnssed when the SR4 channel count rate is 100 eps or when all IIM rnnge switches are above Position 2.

b. IIM Downscale rod block may be typesced when tbr IIM range switch is in the lovest range position.

c. RPM Up enle and ownscale rod blocks my be bypnssed below 30'j rated power.

d. SRM Urncale blocl ray be bypassed whan associate:1 IRM range switches are above Position 6 3 2 / 14 . 2 59 PEI

4 Table 4.2.1 - Continu<4

'tinire. Test and Calibration Frequency For Core Cooline.

Rod 511ock and Isolation Instrueent 7 tion >

In s t rutnent Channel Test (3) Calibration (3) Senser e k ( r.

3. S team Line lov Prersure Note 1 Once/3 mnths rione 1 Ste ra Line High Radi ation Once/ week (5) Note 6 Once/ shift

jPfl ISOIATION

1. Stenn Line High Flow Note 1 Once/3 mnths No e

2. Steam Line High Tem erature Note 1 Once/3 cunths None RC_IC ISOIATION

l. Steam Line High Flow Note 1 Once /3 months None l 2. Steam Line High Tee'arature Note 1 Once/3 months N30c RFACT051 BUlil)ING VENTIIATION

1. Radiation Monitors (Plentm) Note i Once/3 renths Once/shif:

2. Radiation Monitors (Refueling Floor) Note 1 Once/3 months (4' i

'2rF-CAS ISOIATION

1. Rviation Moniters (Air Ejectors) Notes (1.5) . Note 6 Once/sbif:

onst 5

(1) Initia!!y once per nont$ : until exposure hours (M as defined on Figures 4.1.1) is 2.0 x 10 , thereafter accordi?g to Figuro 4.1.1. with ar interval not greater than three renths.

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Table 4.2.1 - Continued e NOTES: ,

I j (2) Calibrate prior to normal shutdown and start-up and theren f ter check once per shift and test once per week until no longer required. Calibration of thi- 1retrinnent prior to normal shutdown means j

a ljustrent of channel trips so that they correspond, v'thir: acceptable range and accuracy, to a '

I simulated signal injected into the instrument (not. pri nary sensor). In addition, II71 gain adjustment

! wi11 be performed, as necessary, in the APPJ1/IIC1 overhp region. ,

l (3) f

! Functionalortests, cperable calibrations and sensor checks nre rot required when the systems are not required to be are tripped. I j to an opernM e status. If tests are missed, thay sh"11 i n perfomed prior to returning the systems ,

(h) Mienever funt handling is in process, a sensor check chall be perfomad once per shift.  !

{i (5) 3 A Functionn1 test of this instrument means the injecticn of a simulnted signal into the instrument (not

- Primary sensor) to verify the proper 'instnmient ch9nnel rasponse clam and/or initiating action.

j (6)

'Ihis inrtrument ref'ueling out agevill beacalibrated with every thrae known radioactive months by wans of a built in currer.t source, and each source.

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LIMITING CONDITiGj5_ TOR OPERATION 4.0 SURVEILLANCE REQUIRDfENTS (i v) The rod block function of the rod worth minimizer shall be verifled by attempting to vi i draw an out-of-sequence con **.1 rod beyond the block point.

(b) Whenever the reactor is in the (b) If the rod worth minimizer is inoper,SJ-startup or run ode below 107, while the reactor is in the startup rr rat ed ther--al pcwer, no control run mode below 10% rated thermal po r rods shall be moved unless the and the second independent operator rod worth minimizer is operable or engineer is being used, he shall or a second independent operator verify that all rod positions are or ongincor verifics that the co rrect prior to comencing withdrawal oporator a t the reactor console or insertion of each rod group.

is following the control rod program. The second operator may be used a" a substitute for an Inope rab i a rod worth minimizer during a ctartup only if tl:e rod worth min nitzer fails a f ter wi t h- l i

d rawal o f a t least twelve control rods.

4. Control rods shall not be withdrawn 4 Prior to control rod withdrawal for for startup or refueling unless at startup or during refueling verify least two source range channels have that at least two source range an observed count rate equal to or channels have an observed count rate greater than three counts per second, of at least three counts per second.

5. Whenever the Engineer, Nuclear, deter- 5. Whenever the Engineer, Nuclear, deter-mines that a limiting control rod mines that a limiting control rod pattern pattern exiets, withdrasal of desig- exists, an instru:nent functional test natad control rods shall be permit tad of the Rkm shall be performed prior to only when the %N system is operable, withdrawal of the designated rod (s) and daily thereafter.

3. 3 / '* . 3 78 RE7

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a j Bases Continued 3.3 and 4.3:

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This is adequate and conservative when compared with the typical time delay of about 210 milli-seconds estimated from scram test results. Approxinately the first 90 milliseconds of the time I l

) Interval results from the sensor and circuit delays; at this point the pilot scram solenoid is deenergized. Approximately 220 m1111seconda later control rod motion is estimated to begin.  !

However, to be conservative, control rod motion is not assumed to start until 200 milliseconds later. This value was included in the transient analyses and is included in the a110wable scram I insertion times of Specifications 3.3.C.1 and 3.3.C.2.

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30 LittITIfF; CONDITIOi!G FOR OPERATION . O mIRVEILIAf;CE REQUIRDIEITTS  ;

i I systece and pump declineralized I water into the reactor vessel.

This test checks explosion of the chsr6e associated with the tested system, proper operation i of the valv s and pu=p capacity. i i Path systems shall be tested

• l and inspected, including each

, explosion valve in the course i

of two operating cycles.  ;

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b. Explode one of two primer assenblies manufactured in the same batch to verify proper function. Then install, as a 5

replacement, the second primer assembly I

in the explosion valve of the syst -n tested for operation.

i i i c. Test that the setting of the system pressure relief valves is between 1350 and lh9 psig.

B. Oparati an with Inapernble Components . Surveillince with Inoperable Camponents Fr -m and af ter the date that a. redun- *'

a hen a co.mponent becomes inoperable, its dnnt c panent is made or found t > be redundant component shall be demonstrated t

to be operable immediately ani daily i j thereafter.

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t sd o on oet sae i i aa t m. t si met ed o nfl e l h rh y r"t hc rra-:

n a a p- l c l e " r ern

, i mul 0 ni gp ct b coioe

'o ooge l: aps c oer lt a m'nih tt e

tf o o i 0 r h sct s a t- t h ne ea r1t -

ehe0 tom"r irpa - " 'u t

oeo - r - - 0t ( u*l r: 1 'pl st h t h ph o h e6n s ul n ie m o yut f ct sh oaegont r t b a

t r2i espa ma f oem ch e sat rs eOs s , f  : t u tO e h t t e h n l

oevo enrrfi n i e .- h1 e t s l otf otii d

l f yrei sgng yDlo reiiernrp n nu c r v w uor atl rt .

t rt th eie nf cat d gt m aro y t g hs lof oe i ein d : ec 1 secyt o- er i rau t t o nnce 1 t r so r '

cnjt nsawi rnae aent nie l td d noe ob ninm o eo d evlb s o ,uee yet vdi ned i e uxsno cncnh s ot il nect  ;

eer rreo qid b nniep d oest np osy oni cgu i q q i oini 7, onn uf pt i ,h ct ot m4 roo amel eb d si t x t ohpt i el ld t aiao art rr1 oc re a a E. e sCt x d ud yot l

ti tf 2 f

a/ r en t

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a f hl i yb b c esnd n ) t h t o .u h . oft i lld d eeenena s e , e rov d q t s o l n namccoco c uncnig eeh t ut e ed ee t erd nt yaa t ood o r t ort or b ct ot n ud ki a siesf p di s st Tocl cui n f o ncu et t e nm r a l v i nof oeqs t a t ss i oe er h e .

pncnia oiodh mo i er e ih ei t cr tl ei e t nh t osaA g s t wdh rt r 5t me u l t n ry e oecoiod n 2ruh l eea 1 o at ero f pt i b rb no 1 el t orb v e,i .hi o e ( s o sa h at el s l ewmb%ai rh pu3 r u t

ti nv l

,)

re et r mct ycu i ocr tl n ti n

t nchk-vu nnrs2 el n e n nmvn u : nt ii a i fi o t r0kl as .

t o eyna mt oeei i t l mf sen ac emo sw hb cmeo0 eobb9on aw f nt airn oe ta sth crncec e e

_ j r o roe irutv.rt t l rr yit n

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d el u e p mch er aig eti a

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s e

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.0 L1:11 TIN I 4'TIhITIOi!O FOR OPERATIOfI h.^ ' "]EILIAIICE B!NUIRD4ENT3 i

4 >

l

4. I be considared operible, the 13 II strn eball nect the following :nnditionc:

i l i i

[ . The IIPCI shall be capable of deliverirs;

[

3,000 riti into the reactor venr.el for ^

l the reactor preseure range of 1120 psig f

l to 150 psig. l 1

! b. The corriensate storage tankr chall

>ntain it least 75,000 g911.ons of j

g udenrate vnter. _

l

! . ne controls for autemtic trunnfer

{ tr the '!i'CI ptep suction from the

.>ndens s te storage tari to the

! napprescion chamber chall be operable.

! [

l h. If the requirements of 3 5.D 1-2 cannat be l met, an orderly reactor shutdown shall be '

I f nit!,tml immediately and the renetor pmnt :m chrtl1 be reduced to 150 psig i vithi n 2h hours thereafter. -

i i

E. Auter;i t i c' I recrure Beller System E. Gurveillance of the Automtic Pressure Relief System shall be perfor.ed as .

follows. f i

f i

e 104 I i

35/h.5

• i i b i

LIMITIIF; COtr?ITIO!C FOR f'PERATION '.O GUIn/EILLAriCE R%UIREMETTf3 30 3 When irradinted fuel is In the reactor vessel and ren-tor coolant tempersture in less than 212 F, all low pressure core and containment cooling nubsystems my be inopernble provided no work is being done which han the potential for draining the reactor ve'isel axcept as al. lowed by spec!fienti.on j.5.G.h below.

h. When 1 rndiateti fuel is in the reactor vesrol und the vessel head is removed, the cuppression chacher may be drained completely and no c: ore than one control rod drive housing or instrument thirble l

opened at any one time provided that l

the cpent fuel pool gates are open and .

l the fuel pool vnter level is maintained l at a level of greater than or eg'lal to 33 feet.

I 21 Deleted 108 REV 3 5lIl* 5

j

.m t

3. 0 LIMITING CONDITNU3 FOR OPERATION h.' SURVEILIAUCE REQUIFEETIS j

i concent mtien in water sh 11 L not exceed reactor coolant temperature 10 min :u ri e ., of total iodine per cc or is above 212 F and analyzed fer water. radioactive iodine of I-131 through I-135. '

i .

(b) Isotopic analysis of reactor cool-l and samples shall be made at Icast i

! 2. . t) Th reactor coolant wat"r shall not once per month. '

! exeed the following limits with '

stortine rates.less than 100,000 2. During startup and at steaming rates

! pennis per hour except e specified below 100,000 pound: per hour, a sanple in 3.6.C.2.b. of reactor coolant shall be taken every Coniuetivity four hours and analyzed for conductivity i

? ):mho/cm and chloride content.

Chloride ion 0.1 ppm (b) For rer.ctor startups the maximum value for conductivity shall not exceed 10 .

>1mho/cm and the maximum value for  ;

l chloride ion concentration shall not

! exceed 0.1 ppm for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> '

l after placing the reactor in the power j oporating condition.

[

'. Except is snacified in 3.6.C.2.b above, the 3 (a) With steaming rates greater than raactor coolant wster shall not exceef the or equal to 100,000 lbs. per hour, s f allowin - limits with sterirg rates creater reactor coolant sample shall ba taken  ;

"nn or equal to 100,000 lig. per hour. at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and when the i Conductivity 10 Arbo/cm continuous conductivity monitorn in-Chloride ion 1.0 prm dicate abne_ mal conductivity (other than short-ttm spikes) and analyzel for conductivity and chloride Ien -

content.

i 3.6/h.6 117 i var

4 j

30 LifilTII!!; CoiTDITie'IS FOR CPERATION .0 Mii&EILI/d:CE PFQtTIRE'ETITS

. b. Prmsurizati :n Te :perature B. Precsuri/ation Temperature

]. "he reactor vessel shall l'e tented e.nd 1. Eeactor vessel shell temperature and r

! pcuer croration shall not l'e conducted renetor coolant pr?ssure shall be unless the reactor vessel tc.mrerature in perreanently recorded whenever the shel]

equal to or grester than trat shown in ter:pertture ic beltv 220 F and the Figure h.6.1. reactor vessel is r ot vented.

The ren"or vessel head bolting stud: chall 2. When the reactor vessel head bolting ntudo not be under tensien unless the tempera- are tightened or loosened, the reactor

+ure of 'he ns vessel head flange and head temperatur" ,

head ar- 570" . chall be per=anently recorded.

3. A neutron flux dosimeter and material samples shall be installed in the reactor vessel adjacent to the vessel  ;

wall at the core midplane level. The material sample program shall con!orm to ASIM E185-66. The neutron f1te. i dosimeter shall be removed during '

the first refueling outage and te sted '

to verify or adjust the calculatei -

values of neutron fluence used te determine the vessel NDTT (Nil Ductility Transition Temperature) from Fig. 4.6.1.

c. O int Chc1 L:try

1. De rem ter c wlant systen r~11onctivity C. Coolant Chemistry

1. (a) A reactor coolant sample shall be taken at least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> when 3.6/h >- 116 RFN P

i  :!"  !} I!,i }!t ji j' jL . j !' [lII l !kiI I (t ,I I:  : .!.

r y rp e re eo , ab .

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r y rmd L i ene I

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su dua e eau V su ur g nt s R ec ra ece y ena _

U h ap x h oe S PV Tvo O Wcm .

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l s l mi t

pl acd eg e r7 sum brnnw e pt c

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y eb rauo t rurh seaiy e r arep rf adt t n httb ur r ul meet somei i dt na D

sl ipvd sar o e t

nnr i rn n id o or% h a r eh pnrs e i nct 5 e rs opeh ri n r N b o e ieo bl ei ht e bt siif i

e prhca een 2o O f m - sh ymc t ahl w i ot nh I a l rt na rtl rc rt ot e T h l e ah e ft anoe uc ,c nd A C ek n ch osnt sgc s ,h et p oan s eo R waed t hewns m E n yeh nn n nou ya P o rrwaon o ompexe s roee O db i et rora

- d gl on i

s s ,ii s

i t td ue et y cu R hh uxo r O

s nredsd a e oh%et 2thot y F

es r r eemeee r l 1 - p5h p s e pe t bi rrr mti pu t

n pnl s maaon t e c ee h ngre ed .h t

S pk ea upc e o cmarx ti nec g N ua hhl que c cmith ee cih u ng O Sn t cl ess n a rat v nawpd n I

T

- ar o r rt e , il os e i .

I aP f n ee C egct sn e h pl or rv:

D r

um oot srb i aiu n t oensed f rl eeh o t l m aw eo N su t s sl e Apeml wm ioot e O su hset sl g Wtf ab l d C ec gel nea y ra i rb erh x . .

G p 'l Epamps O a b N

I T .

I L 5 M

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0 7 3 3 h , ;l  :! ill!)]1{ ;j1 1{< I; j j'  :! l J;. 4

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3.o LUtIT1UG CONDTTIOrn mR OPERATION h.r. SU; VEIIHdICE 10%UIPDINTS

.s. If fhe sp+ 'ifications of 3.7. A cannot be mot, the rmetor shull be pliced in a '

cold chutdmT) condition within Ph hours.

p. Standby i;as Treatment System H. Standby Cas Treatment System

1. Except as cycified in 3.7.B.3 triow.

both circui+n of the standby cts treit- 1. Standby gas treatment system curveillance cent syste~ rhall be operable at 'ill shall be performed as indicated below:

i + !mer when second try contain-ent decttty in required

• a. At least once per operating cycle it i shall be demonstrated that:

(1) Pressure drop across the combined high-efficiency and charcoal filters is less than 7.0 inches of water, and (2) Tnlet heater output is at least 15 kw.

! b. Within 30 days of the beginning of ench r<>rueling outage, whene er a filter in change

whenever work is performed that could affect filter systems efficiency, and at intervals not to exceed six months between refueling outages, it shall be demonstrated that:

(1) The removal efficiency of the installed particulate filters for particles having a mean diameter of 0.7 microns shall be 3 eh.7 lh8 BEJ i

i

l t

a l i l

3.0 1.1 M111;'G CONDITIN S FOR OPERATIO" 4.0 Sin?"EILIA!!CE REQUIREMENTS C. Dcondary contai, ment C. Secondary Contninment i 1. Second,ry containment integrity, shall be

1. Secondary containment surveillance shall naintained during all modes of plant be perfomed as indicated below:

l operation except when all of the following conditions are met.

3 7 The reactor is suberit ical and Specifi- n. Secondary containment capability to

cation 1.3.A is met. maintain at least a 1/4 inch of water i vacuum under calm wind (2 < u < 5 mph) conditions with a filter train flow rate of64,000 scfm, shall be dem-onstrated at each refueling outage l prior to refueling. This surveillance j testing should be reported in the

b. The reactor water temperature is below semiannual operating reports. ,

I 212 and the reactor coolant system is '

vented.

j

! I

c. i:o activity is being performed which t can reduce the shutdown margin below ,

that spe-I fied in Specificat ion 3.3. A. i j

1.7/4.7 150 REV i

,l~ >  ! i5 if +  ! tt: !Lj ; * [i k !! [L [ i}  !(!  !:(!f ;

e .

s

d . r es e w e t e e p de y

_ n o el h am h o eh l t li t wr 1 7 vl l o ere ut c el op ttMl a

rrtn a 5 E 1 R af l a yi e lfr eoa v s c l sr c o pfre na ytl u yef o eb n o caars cnid - t p e hcoo rpol v id gtifl n

gl e ee l t e t c n t wcra l

a am isad it s r oxeh .

V l r t emed t ne e pewsd oo avot n aet t ( op) n e n

sf rlt sa eaue rm eue r

a n

es peeep se 1 r o e pvat n prb u i

t t p n o o ot q ovnvo a ndei sl lil e ee ronbt rnl r yal ar S

l mb eia a eia e l v v o n pt li p h p l m d I

s i l al adlt el eai ee ms e anann moeoa M t t a nh coth n nsasi ct s nse c

n riti ot std D n os oir oyv . o na ae R e c edc sl n l nl s I m e t epei t ao t loioo U n yc sl ott syvi s lsasl Q i rn ab aa ar t a Aimic E

R a aa earim eak a e t ml l ret o l mcr l n il ewit iee )

E o ri t ponu t rh p t 1 C C pe A o pi a Apcc A (

. sr y er v

L r hu . . .

L a Ts a b c I m E i V r .

R P 1 U

S I'

0 k

1 t p

,7 e n e c i r d3ex o o nc lt mei ec l l e ua nb l f e uatb r n rTnaer .

d e ee t5t t v enme hi up at 1 t ro i V dt T

I 0

dni i n nese it nb I rh o si .

T rt i si l2 A ii t il ml R w a E

P r

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F c am t n

evyvn r rli g e n a C

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n eomvd hii e O uei i t t rki

. I fbd a apcr l t rl "i R' "ti n eo1h c

. D T

h ou o vs1 ce I Tnb C ei- 1 p O n ws C c

' eld o

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,il . i-! :i ,ij, ;jl! j . ij{j 1 - -

e 3.0 LIMITI!M COI" ITICIG FOR GPERATIO:7 h . r- SUM'EILI/JCE FIQUIREIMS

c. At least once per quarter - Continued (2) With the reactor power less than 75% or rated, trip main stcv isolation valves (one at a *.Inc) and veri fy closure time.

d. At least once per week the main stca.9-line power-cperated isolation val 70s shall be exercised by partial closurt ard subsequent reopening.

2 In the ment any isolation valve 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 operation in the run mode may continue at least one fbily closed valve in each line provided at least one valve in each line having an inoperable valve shall be recordad having an inoperable valve is closed. daily.

3. If Specifica tion 3. 7.D.1 and 3. 7. D. 2 cannot l be met, initiate nonnal orderly shutdown I and have reactor in the cold shutdown l condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

, ,, : 152

.1<h...i py.;

e ,

n 9

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SI'RVEILIAJCE REOUIREMENTS s

6. The release rates of radioactive 1. Gaseous release of tritfum shall be particulate- with half-lives greater calculated on a quarterly basis frot than 8 days shall not exceed 8 percent tritium concentra-ion of the condensate.  !

, of th.= limit in Specification 3.8.A. Vaporous tritium ; hall 12 calcalated ave ra ged over any calendar quarter. from a representative auple. The sum of these tro values shall be re-

7. If the maxirum release rate limi t s of Spec- ported as the total critiu= release.

ifications 3.8.A.1, 3.8.A.3, or 3.8.A.5 are not mt following a routine surveillance 4. Radiciodine and radioactive particulatas with l check, an orderly shutdown shall be half lives s reater than 8 days released from .

initiated ami the reactor shall be in the the off-ga. *ck and reactor building vent

• i I

cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. shall be er *ously sampled. Station record of release il radiciodine 131 and partic-

8. If the Ifmits of Specification 3.8.A.2, ulates witt '.f lives greater than 8 days 1.8.A.4, or 3.8.A.6 are exceeded, shall bc x f r.ained on tie basis of all stack appropriate corrective action such as an and vent cartridges counted. Ihe charcoal ordarly reduction of power shall be cartridges shall be counted weekly when the j initiated to bring the releases within measured release rate of radioiodine 131

! these limits. activity is less than the limit of Specifi-

' cation 3.8.A.4; otherwise stack cartridges

9. If the relene rates exceed four percent shall be counted daily if the stack I-131

! of the limits in Specification 3.8.A.1 contribution exceeds 507. of the limit of averaged over any calendar quarter or two Specification 3.8. A.4 and vent cartridges percent of the limits in Specifications shall be counted daily if the vent I-131 l 3.8. A.3 or 3.8. A.5 averaged over any contribution exceeds 507. if Specification

calendar quarter, the following actions 3.8.A.4. The particulate filters shall be shall be taken

counted weekly when the measured release rate of particulate radioactivity with half-1 lives greater than 8 days is less than the limit of Specification 3.8. A.6; otherwise i l,

stack filters shall be counted daily if the stack particulate contribution exceeds 507.

of the limit of Specification 3.8. A.6 and vent filters counted daily of the vent l particulate contribution exceeds 307 of the 3.P/4.3 limit of Specification 3.8.A.6. .

i 170 REV

9

).

i 30 LIMITITM c?.'DITI"fl3 FOR OFFRATIDIT h.0 SURVEILIMCE PEOUIRB4ETITS

c. J sotopic analyses including detemination of tritium of representative ba :hes of liquid effluent shall be perfomed and /

recorded at least once per quar-tar. Each batch of effluent relassed chall be counted for gross alpha and beta activity and the results recorded. At least once per month a gm::t:n

. can of representative batches

- T effluent shall 1.e perfomed

• nd recordad to detemine the gar:rn energy peaks of these batches. If energy peaks other  ;

than those detemined by the "

Frevious isotopic analyses are found, a new set of isotopic analyses shall be perfomed and  ;

recorded.

i

d. Delete,

e. The liquid effluent radiation  ;

m.>nitor shall be calibrated r q ua rterly.

172 3.8/4 8 ,

m

o t

I l

1 i

! 3.0 LIMITING CONDITIUN FOR OPERATION I.0 4

SURVEILLAIICI. REQUIREtBITO

! P. Both diesel generators are operable and j capable or feNing their denjenatM bl60

! volt busos.

j i 3 A second nource of off-site power ( reserve r i transfor er lAR) in fully operational and

} l energize ! to carry power to tha plant l hl60V ac busen.

• 1  !

4. (a) hl60V Buses #15 and #16 a re energized.

4 ( b) 480V Load Centers #103 qwl floh are  !

energized. f i

l S. All statien 24/48,125, and _2';0 volt j

batterien are charged and in service, and

) associatnl battery chargers are operable. i P

1 t n. k' hen the mode switch is in Run, the avail- '

! abili'y of electric power shall be as spec' l

ified in 3.9. A, except as specified in 3.9.B.1,  !

3.9.B.2, 3.9. B.3 and 3.9.B.4 or the reactor l ,__

shall be placed in the cold shutdown condition  !

l within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,  !

j 1. Transmission Lines '

1 From and after the date that incoming i,

power is availabic from only one line,

, reactor operation is permissible only j during the succeeding seven dafs unless  ;

an additional line is sooner pinced in 181 i 3 9/h.9 FEV 2

i

o 4

s 3.0 LIMITIrr; CONDITION 3 FOR OPERATION 4.0 GURVEILIl&'E I'F4UIRD4DTS

c. For the diesel genentors to be c. During each refuellrw; outage, considered cremble, thare shall be a minimum of 26,250 gall' ens of diesel the conditim.c under which the diesel gmentors are required t

fuel (7 days supply for 1 diesel gen- will be simulated and tests cr n- ,'  ;

er, tor at full laad) in the diesel '

ducted to denmstr,te that they oil stomge tank.

will start and be ready to accept the '

m rgency load within ten seconds.

d. During the monthly generutor test, the diesel fuel oil transfer pump i and diesel oil service purp shall be opernted.

e. Once a month the quantity of diesel fuel available shall be logged.

f. Once a month a cample of diesel fuel shall be taken and checked for quality.

i 4 Station Battery Syste= * "

• f If one of the two 125 V battery systems or the a. Every week the specific gravity '

i 250 V battery system is made or found to be and voltage of the pilot cell i

inoperable for any reason, an orderly shut- and te=perature of adjacent cells down of the reactor will be initiated and and overall battery voltage shall the reactor water temperature shall be be censured.

reduced to less than 212 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless such battery systems are sooner made operable.

3.9/4 i 183 -

\ Psi (

t

, AEC DIST RILUTION FOR PART 50 DDCKET MAT ERI AL l (ll; , DR ARY f OH'/)

CON 1RUL NO 12033 FILE- l F ROM: Northern States Power DATE OF DOC DATE REC'D LTR TWX RPT OTHER Minneapolis, Minn. 55401 Mrm lub_Mayer 11-15-14 11-26-14 Y l

TO: ORIG CC OTHER SENT AEC PDR XXX j

E.G. Cape 3 signed XXX SENT LOCAL PDR ~

j CTASS UNC L ASS PROPINFO INPUT NO CYS REC'D DOCKET NO: l XXX XXX 40 50-263 DESCRIPTION: ENCLOSURES:

L2r requenting an andt to the operating -

Proposed changes to the tech specs. . . . .

Aicense....trans the following.... notarized consist of changes to the tables of contents 11-15-74...... and minor changes due to typographical errors......

ACKNOWLEDGED (40 cys enc 1 reed)

PLANT N AME:

Menticelle FOR ACTION /lNFORMATION 11-26-74 JE E L'T L E P ( L W ..E:.CEF iL)eCIE' W 's (L) REG AN (E) w Com W Coms WeConss W, Conies CLARK (L) STOLZ (L) DICKER (E) LE AR (L) nn NOT REMOVE

.L> U W/ Copies W/ Copies W/ Copies W/ Copies PAR R (L) VASS ALLO (L) KNIGHTON (E)

W/ Copies W/ Copies W/ Copies W/ Copics KNIEL (L) PURPLE (L) YOUNGBLOOD (E)

W/ Copiu W/ Copies W/ Copies W/ Copics INTERNAL DISTRIBUTION (f41EG FC TECH REVIEW DENTO3 L IC ASST A/T IN D VAECfDR G R i f.1 E S B R AIT *.Z

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