ML20073P950

Amend 17 to License DPR-22,revising Tech Specs to Change Title from AEC to Commission,Correcting Table Numbering & Clarifying Bases Section to Reflect Removal of Two Vacuum Breakers

ML20073P950
Person / Time
Site:	Monticello
Issue date:	04/17/1983
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Northern States Power Co
Shared Package
ML20073P951	List: ML20073P950 ML20073P952 ML20073P955
References
DPR-22-A-017 NUDOCS 8304270329
Download: ML20073P950 (13)

Text

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. UNITED STATES 8

NUCLEAR REGULATORY COMMISSION g

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NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.17 License No. DPR-22 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Northern States Power Company (the licensee) dated September 24, 1982 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this $mendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements

.have been satisfied.

l 2.

Accordingly, the license is amended by changes to the Technhal Specifi-l cations as indicated in the attachment to this license amendment, and

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paragraph 2.C.2 of Facility Operating License No. DPR-22 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No.17 are hereby incorporated in the l

license. The licensee shall operate the facility in accordance with the Technical Specifications.

83042703298304[g DR ADOCK 05000263 PDR

2 3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY C0tMISSION

/60 Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: April 17,1983 O

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s ATTACHMENT TO LICENSE AMENDMENT NO.17 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT l

4 4

90 90 126 126 150 150 179 179 180 180

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227c 227c 230 230 241 241 253a 253a l

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e e #8 e

Is.

protective Function - A. system protective action which results from the protective action of the channels monitoring a particular plant condition.

R.

IE70 thernal megawatts. Rated Neutron Flux - Rated flux is the neutron flux that correspouds to a steady-state power level of 8.

Rated Thermal Power - Rated thermal power means a steady-state power level of 1670 thermal megawatts.

T.

Rep.ctor Coolant System pressure or Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical 8pecifications are those existing in the vessel steam space.

U.

Refueling Operation and Refugling Outage - Refbeling Operation is any operation when the reactor water temperature is less than 212F and movement of fuel or core components is in progress. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a' regularly scheduled refueling outage; however, where such outages occur within 8 months of the completion of the previous reibeling outage, the required surveillance testing need not be performed until the next regularly scheduled outege.

t V.

Safety Limit - tha. si.fety limits are limits below which the maintenance of the claddire and primary system integrity are assured. Exceeding such a limit is cause for plant shutdown and review by the l

/

' /

Cceanissida before restusption of plant operation.

Operation beyond such a limit may not in itself l

i result in serious consoquences but it indicates an operational deficiency subject to regulatory review'.

W.

Seconde.ry Contaisusent Integrity - Secondary Containment Integrity means that the reactor building.is closed,and the following conditions are metts i

4 1.

At least one door in each access opening,is closed.

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f. The standby gas treatment system is operable.

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3., All reactor building'ventilahlon system automatic isolation valves are operable or are secured y

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in thes, closed position.

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X. 7 easor Check - A qualitative determination of operability by observation of sensor behavior during 8

f ' operation.

This determination shall inclode, where possible, comparison with other independent

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seneors measuring the seine variable.

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L 1.0-i Amendment No. 17 4

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i seene centinned 3.3 and 4.3:

%e analysis assumies 50 milliseconds.for teactor Protect ion system delay, 200 milli seconde from de-energizat isme of scree solenoids to alie beginning of rod motion, and 175 millisecos Je later slee ends are at the SI posit iosi.

Section 3.3.C.3 allows a lower HCFR limit to be used if the cycle average scram time (Tms ). le less these the miljiested analysle menia scram time ("Is) (see Neference 7, of section 3.11)

% is the welghted cycle average scram time to the 20! Insertion poeltlen (~ notch 34) of all the operable rode seessured at any point in glie cycle.

the nimmber of surveillance teste performed Meeres n a to date In thle cycle.

T l_

N T g I Ng. number of cont rol rode meseared in the I"I ith test.

'l'i sverage scrasi t ime.to alie 201 insert fun

=

N poelt los of all roJe measured la the i th 1*l test.

T, le the adjusted analysis mean screas t ime diere Ng = total nasuber of active roJe measured les to the 201 lasertion poeltion.

the first test following core alt erat ions.

0.750 = the mean scram time used in tlie g

analysis.

a N

0.0875 = 1.65 0.053 ideere 1.65 is the appropriate i

g f

T, = 0.710 4 0.0875 st atist ical niumber to provide a 951 l

conflJence level and, 0.053 la the k

N J

st anJard devlet ion of alie dist ribut i.no

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for everage scram insertion time to the 208 peelt lon, tlant was insed In the

.4 analysis.

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t 3.3/4.3 BASES 90 1

l Amendment No. 5 17 l

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30 LIMITING CONDITIONS FOR OPERATION k.0 SURVEILLANCE REQUIREMEN1B D.

Coolant Imakage D.

(bolant Ieakage

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

vessel and coolant tempe nture is above 212 F Any time irradiated fuel is in the nector 0

reactor coolant system leakage, based on vessel and coolant temperatum la above s

sump monitoring, shall be limited to:

2120F, the following surveillance y mgram shall be carried out:

a. 5 gym Unidentified Leakage

b. 2 gym increase in Unidentified

a. Unidentified and Identified Leakage rates shall Leakage within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Period be recorded at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> using

c. 20 gym Identified Leakage primary containment floor and equipment drain t

sump monitoring equipment.

d. no pressure boundary leakage

2. With reactor coolant system leakage greater

b. Primary containment atmospheric particulate radioactivity shall be recorded at leaYt" ~

than 3.6.D.1.a or 3.6.D.1.c above, reduce the once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

l leakage rate to within acceptable limits within four hours or initiate an orderly shutdown of

c. Drywell pressure and temperature shall be re-the reactor.and reduce reactor water tem corded'at'least osce every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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i ture to less than 2120F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.pera-2.

The reactor coolant system leakage detection l

3. With an increase in Unidentified' Leakage in ex-systems shall be demonstrated OPERABLE by:

cess of the rate specified in 3.6.D.1.b, ident-l ify the source of increased leakage within four

a. Primary containment atmosphere particulate houts or initiate an orderly shutdown of the monitoring systems-performance of a sensor check at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a channel u

reactor and reduce reactor kater temperature to functional test at least monthly and a less than 2120F within '24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

channel calibration at least once per cycle.

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

.b. Primary containment sump leakage measurement d

and 3.6.D.3 above are taken,. initiate an order-system-performance of a sensor check at l

ly shutdown of the reac. tor and reduce reactor least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and a channel calibra-water temperature to less than 2120F within 24 tion test at least once per cycle.

hours.

5. At least one of the leakage measurement instru-i l

ments associated with each sump shall be opera-ble and the drywell particulate radioactivity monitoring system shall be operable or a sample of the containment atmosphere shall be taken i

and analysed at least every four hours. Other-i I

wise, initiate an orderly shutdown of the reac-ter and reduce reactor water temperature to Amendment No. U,17 i

less han 2120F within 24 hars.

3.6/4.6 126 i

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3.6 and 4.6:

anses Continued l

4 D.

Coolant Isakage I

h anowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentany observed behavior of cracks in pipes. Se nomally expect.ed background leakage due j

to equipment design and the detection capability of the instrumentation for detemining leakage was 1

i also considemd. he evidence obtained from experiments suggests that for leakage somewhat greater j

than that specified for unidentified-leakage, the pmbability is

===11 that the layerfection or crack associated with such leakage would gmw rapidly However, in an cases, if the leakage astes exceed 4

the values specified or the leakage is located and known to be Pressure Moundary 1.aakage and they cannot be re-duced within the allowed times, the reactor will be shutdown to allow further invescigacion and corrective

'setion.

I 20 leakage conection sumps are provided inside primary containment. Identified leakage is piped

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from the recirculation pump seals, valve stem leak-offs, reactor vessel nange leak-off, bulkhead

.and bellows drains, and vent cooler drains to the dryven equipment-drain sue. An other leakage is conected in the dryven floor drain sump. Both sumps are equipped with level and now trans-l l

' mitters connected to recorders in the cmtrol room. An annunciator and computer alam are pm-4

, vided in the control room to alert operstors when anoweble leak rates are approached.

Dryven

' airborne particulate radioactivity is continuously monitored as won as drywell atmospheric tem-perature and pressure. Systems connected to the reactor coolant system boundary are also monitored for leakage by the Process Liquid Radiation Monitoring System.

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he sensitivity of the sump leakage detection systems for detection of leak rete changes is better i

i than one gym in a one bour period. Other leskage detection methode provide warning of abnomal Isakage and are not directly calibrated to provide leak rate measurements.

E.

Safety / Relief Valves j

l Testing of an required safety / relief valves -each refueling outage ensures that any valve deterioration is detected.

j A tolerance value of 1% for safety / relief valve setpoints is specified in Section III of the ASME Coller 1'

and Pressure Vessel Code. Analyses have been perfomed with an valves assumed set 1$ higher (1108 psig

+ 1%) than the nominal setpoint; the 1375 pois code limit is not excseded in any case.

i h safety / relief valves are used to limit reactor vessel overpressure and fuel thermal duty.

h required safett/ relief valve steam now capacity is determined by analyzing the transient accompanying

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the mainsteam.now stoppage resulting from a postulated M11V closure from a power of 1670 Mi. h analysis g

assumes a multiple-failure wherein direct scram (valve sition)isneglected. Scram is assumed to be fran l

indirect means (high flux). In. this event, the safety lier valve capacity is assumed to be 83.2% of the full power steam generation rate.

3 6/4.6 BASES Amendment No. M 17 H0

gases Continued:

One-inch opening of any one valve or a 1/8-inch ope'ni'ng for all eight valves, measured at the bottom of the.Jise with the top of the disc at the seat. The position indication systen as designeJ to detect closure within 1/8 luch at the bottom of the disc.

each refueling outage and following any sigificant maintenance on the vacune breaker valves, At positive seating of the vacuum breakers wit! be verified by leak test.

The leak test is conservatively designed to demonstrate that leakage is less than that equivalent to leakage through a one-inch i

orifice which is about 31 of the maalmum allowable. This test is planned to estabitsh a baseline for j

valve performance at the start of each operating. cycle and to ensure that vacuum breakers are maintained as nearly as possible to.their design condition.

condition for operation.

This test is not planned to serve as a 11afting

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During reactor operation, an exercise test of the vacuum breakera will be conducted monthly:

test will verify that disc travel is unobstructed and allt provide verification that the valves are This closing fully through the position Indication system. If one or mote of the-vacuum breakers do not 4

seat fully as determined from the indicating system, a leak tegt will be conducted to verify that leakage is within the maximum allowable.

is approutmately 1/16". the planned test is designed to strike a balance between the detection switch capability tn verify closure and the maximum allowable leak rate.

establish the basis for this limiting condition.

A special test was performed to breakers were shimmed 1/16" open at the bottom of the disc.During the first refueling outage all ten vacuum The bypass area associated with the shimming corresponded to 631 of the maximum allowable.I The results of this

3. 7.1.

Two of the original ten vacuum breakers have since been removed. test are shown in Figure 1

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When a drywell-suppression chamber vacuum breaker valve is exercised through an opening-closing cycle the position indicating lights at the remote test penets are designed to function as followas Full Closed 2 Creen - On 2 Red

- Off*

Intermediate Position 2 Creen - Off 2 Red

- Off i

Fut! Upen 2 Creen - Off i

l 2 Red

- On l

The remote test panel consists of a push button to actuate the air cylinder for testing, two red lights, l

179 3.7 BASES Amendment'No.17

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Bases continued:

and two green lights for each of the eight valves. There are four independent limit switches on each valve. The two switches contrutting the green lights are adjuse.e1 to provide an Indicattan of dise.

openin' of less than 1/R** at the bottom of the disc. These switches are also used to activate the s

valve position alarm clernits. The tuo switches controlling the. red lights are sdjusted to provide indicatien of the disc very near the fuit open position.

the control room starm circuits are redundant and tatt safe. This assures that no simple f ailure will defeat alarming to the control room when a valve is open beyond allowable and when power to the switches l

fails. The alare is needed to alert the operator that action must be taken to correct a malfunctica or to investigste possible changes in valve position status, or both. If the alare cannot be cleared due to the inability to estebitsh indication of closure of one or more valves, additional testing is required.

The alarm system allows the, operator to make this evaluation on a timely basis. The frequency df the testing of the alarms to the same as that. required for the position Indication system.

Operablitty of a vecuum breaker valve and the four essociated indicating light circuits shall be estabitehed by cycling the valve. The sequence of the indicating Ilghts will be observed to be that previously described. If both green light circuits are Inoperable, the valve shall be conside~ red inoperable and a pressure test is required issuediately and upon indication of subsequsnt operattan, if both red light circuits are looperable, the valve shall be conside' red inoperable, however, no pressure test is required if positive closure indication is present.-

i The SI osygen concentration minimises the poselbility of hydrogen combustion following a lose of coolant accident. Signif tcant quantitles of hydrogen could be generated if the core cooling systems failed to sufficiently cool the core. The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown to more probable than the occurrence of the loss of coolant accident upon which the specified oxygen enneentration limit is based. permittles access to the dryvell for lesk Inspections during a startup is judged prudent in terna of the added plant safety offered without significantly reducing the margin of safety. Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, uhen the primary system is at or near i

rated operating temperature and pressura. The 24-hour period to'grovide inerting is judged to be sufficient i

to perform the leak inspection and estabitsh the required oxygen concentration. the primary contatsusent is normally slightly pressurised during Periods of reactor operation. littrogen used for toerting could leak out of the containment but air could not teak in to increase oxygen concentration. Once the con-tainment is fitted with nitrogen to the required concentrations no monitoring of oxygen concentration is necessary. Ilowever, at least once a week the oxygen concentration will be determined as added assurance.

3.7 BASES s 130 Amendment.No. 17

i TABLE 3.13.1 4

SAFETY RELATED FIRE DETECTION INSTRUMENTB Minimum Instrumenta Operable 1

Fire Zone Location Heat Flame Booke IA "B" RXR Room 3

IB "A" RHR Room 3

j IC RCIC Room 3

lE HFCI Room 2

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

23 Reactor Bids. 935'. elev - CRD NCU Area East 10 l

2C Reactor 51d. 935' elev - CRD HCU Area West 11

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3 2E Reactor tids. 935a LPCI Injection Valve Ares t

l 35 Beactor Bldg. 962' elev - BBLC Area 2

.1C Reactor Bldg. 962' elev - Bouth 5

3D Reactor Bids. 962' elev - BBCCW Pump Area

.4

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4A Reactor Bids. 985'.elev - Bouth' 4

45 Reactor Bldg. 985' elev - RBCCW Hz Area 5

4D BBCT Byaten Room 2

5A Reactor Bldg. 1001' elev - South 7

5B Reactor Bldg. 1001' elev - North 3

SC Reactor Bids. - Fuel Foot Cooling Pump Area 1

I 6

Reactor Building 1027' elev 5

7A Battery Room 1

73 Battery Room 1

7C Battery Room 1

8 Cable Spreading Room 7

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

13C Turbine Bids. - 911' elev - HCC 133 Area t

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

l 15A

1. 12 DC Room & Day Tank Room 3

I 155 fil DC Room & Day Tank Room 3

16 -

Turbine Bldg. 931' elev - Cable corridor 3

j 17 Turbine Bldg. 941' elev - Cable Corridor 3

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

.5 195 Turbine Bldg. 931' elev - HCC 142-143 Area 1

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

i 20 Heating Boiler Room I

23A Intake Structure Pump Room 3

3.13/4.13 227c Amendment No. 17

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5.0 DESICH FEATURES 5.1 Site f

l A.

The reactor center line is located at approximately 850,810 fact North and 2,038,920 feet East as determined on the Hlonesota State Grid, South Zone. The nearest site boundary is approximately 1630 feet S 30* W of the reactor center line and the exclusion area is defined by the sinteum fenced area shoun in FSAR Figure 2.2.2a.

Due to the prevailing wind pattern,*the direction of l

manimum integrated dosage is SSE. The southern property line follows the northern boundary of the right-of-way for the Bur 11ogton Northern Railway.

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5.2 Beactor -

j A. The reactor core shall consist of not more than 484 fuel assemblies.

B. The reactor core shall contain 121 cruciform-shaped control roda. The control rod material shall j

be boron carbide powder (5 C) compacted to approximately 70% of theoretical density.

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l 5.3 Reactor Vesset

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l A. The pressure vesset shall be designed for a pressure of 1250 pois and a temperature of 562* F._

j The coolant recirculation system shall be designed for a pressure of 1848 pais on suction side of pump and 1248 pois at pump discharge.

The applicable design codes shall be as described in Sections 4.2.3 and 4.3.1 of the Monticello Final Safety Analysis Report.

54 containment A. The pr' mary containment shall be of the pressure suppression type having a drywell and an gbsorption i

chamber constructed of steel. The drywell shall have a volume of approminately 134,200 ft and is designed to conform to ASHE Boiler and' Pressure Vessel Code Section til Class S for an internal pressure of 56 psig at 281 F and an external pressure of 2 ps at 281 F.

The absorption chamber shall have a total volume of approutmately 176,250 ft i

5.0 230 m

l Amendmerit No.

17 f

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O eratlone Committee (DC)

B.

P l.

Membership He Operations committee shall conslet of at least six (6) members 4rsun from the key super-visors of the on-site supervisory staf f.

H e FIsnt Hausger sisell serve as Qialtman of the DC and shall appoint a Vice Chairman from the DC memberely to act la ble absence.

2.

Heating Frequency n e operations committee will meet on call by the Qialream or as requested by'Individaal members and at least isonthly.

3.

quorum A quorum shall include a majority of the permanent members, including the Qialtman or Vice Osaltman 4.

Responsibilities - The following subjects steelt be reviewed by the operations committees a.

Proposed tests and esperleents.and their results.

b.

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

c.

Proposals whlcis would effect permanent, changes to normal and emergency operating procedures!and any other proposed changes or procedures that are determined by

.the Plant Hansger to af fect nuclear safety.

d.

Proposed changes to the Techalcal Speelfications or operating licenee.

i e.

All reported or suspected violatione of Technical Specifications, operating Ilcense requirement s administrative procedures, or operating proc,edures. Results of Investi-get lons, including evalust lon and recommendet lons to prevent recurrence, will be reported, in writing, to the General Manager Nuclear plants and to time Qistrean i

. of the Safety Audit Cosimitt ee.,

6.2 241

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Amendment No. $ 17 l

4 3.

Special Reports 4

When radioactivity levels in samples exceed limits specified in Table 4.16.3 a Special Report shall he submLtted within 30 days from the end of the af fected calendar quarter. For certain cases 3

involving long analysis time, determination of quarterly averages may extend beyond the 30 day period.

In these cases the potential for exceeding the quarterly limits will be reported within the 30 day period to be followed by the Special Report as soon as practicable.

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Amendment No. J6 17

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