ML20148G077

Amend 21 to License DPR-3,incorporating Util Core Performance Analysis,Changing Operating Limits & Adding Requirements for Compliance w/10CFR50.46 Provisions,Thus Terminating 741224 Order Limitations

ML20148G077
Person / Time
Site:	Yankee Rowe
Issue date:	12/05/1975
From:	Goller K Office of Nuclear Reactor Regulation
To:
Shared Package
ML20148G075	List: ML20148G072 ML20148G077 ML20148G085 ML20148G092 ML20148G096 ML20148G102
References
NUDOCS 8011070166
Download: ML20148G077 (14)

Text

- _ _ _ _ _ _ _ _

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k UNITED ' STATES.

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- NUCLEAR ' R5GULATORY COMMISSION '

E WASHINGTON. D.

C. 2055E

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YANKEE ATOMIC ELECTRIC COMPANY 2.:-

-DOCKET NO. 50 -YANKEE NUCLE R POWER STATION (YANKEE-RONE)

[h

AMENDMENT TO FACILITY OPERATING LICENSE h:2 Amendment ho. 21

==

License No. DPR-3' 1.

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

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=.

A.

The applications for amendment by Yankee Atomic Electric

. Company (the licensee) dated July 14, 1975 [as supplemented

==

October 10, October 28, November 7 (Proprietary Information appended), November 21, and November 26, 1975); July 8, 1975 (as supplemented November 24, aM November 26,1975); and September 23, 1975, cornply with the standards and require-

~~

monts of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I;

~ ~ ;:3 3

B.

The facility will operate.in conformity with the applications, '

s!

the provisions of the Act, and the rules and regulations of

!j the Commission; "9

a C.

There is reasonable assurance (i) that the activitics authorized g

by this amendment can be conducted without endangering the

)lj health and safety of the public and (ii) that such activities nj will be conducted in compliance with the Commission's regula-li tions; and y

D.

The issuance of this amendment will not be inimical to the common 3

defense and security or to the health and safety of the public.

2.

Accordingly, Facility License No. DPR-3 is hereby amended by revising

- 5, Paragraph 3. A. (2) and adding Paragraph 3. A. (3) as follows:

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"(2) Technical Specifications

3

9.

The Technical Specifications contained in Appendix A, v'"1 as revised, are.hereby incorporated in the license.

The licensee shall operate the facility in accordance 3

with the Technical Specifications, as revised by issued

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changes thereto through Change No.y 2 8

" (3) Yankee shall not operate the facility with Core XII beyond low power physics testing until the modifica-tions to precludo single failures in the ECCS (as

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described in the staff Safety Evaluation issued with Amendment No.2 1) have been completed.

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3.

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

P O b5 FOR Tile NUCLEAR REGULATORY C0hNISSION Odgr al F "ned by

~

l Karl Galler Karl R. Goller, Assistant Director for Operating Reactors Division of Reactor Licensing

Attachment:

Change No.12 0to ihe li Technical Specifications Date of Issuance:

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- Ji.i NITAC191ENT TO' LICENSE AMENDMENT' NO. 2'1 E E=!;#;

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-Cl!ANGE NO..t 2 6'TO TIIE TECl!NICAL SPECIFICATIONS'

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FACILITY OPERATING LICENSE NO. DPR-3 r...

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DOCKET NO. 50-29 si!

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' Revise ~ Appendix A as follows:

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Reraove Pages -

Itisert New Pages

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j-1 5b (Figure 8-1)

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Tc,rm Ar.0-318 (Rev. 9 53) AICLt 0240 1lr ua sa novsaniaswT Painviae orricas sen.sas.ies t

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. -., - -. -, - -,, - -.. - ~... +. -...

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~ 21h Nuclear Instrumentation System

_2.15 Radiation Monitoring System M

218 Fuel llandling System 22h -Compressed. Air Systems 231 Vapor Containment

,-232 Radiation Shielding

~

- 235 Archetectural Features Physical. arrangements of structures and equipment will be as

. described in' Section 200 of the license' application. Mechanical equipment and systems 5d.11 be interconnected as shown in'the Fundamental Flow Diagram included in that section.

1

-Electrical equipment and systems which provide station auxiliary

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. power supply will be as described in Section.226 of the' license 9

application' and trill be interconnected as shown in the 2h00 volt one-line diagram and the h80 volt one-line diagram, sheets 1, 2 and.3, included in that section.

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The ventilation system for the control room area, radiochemical p;

laboratory, decontamination cubicle, fuel transfer pit house, and other potentially contaminated portions of the Turbine Generator Service, Primary Auxiliary, and Waste Disposal Buildings shall be-in accordance with the description contained in Section 228 of Part B of the license application.

~

3.

The Performance Analysis for the current reload core, " Yankee l

Nuclear Power Station Core XII Performance Analysis, July 9, 1975" (as supplemented October 10 and November 26, 1975), is j

' incorporated,as part of these technical speciitcations. The f 26' J

analysis presented in the FSAR for Core XI forms the basis for d

the reference core performance analysis.

M b"

C.

PDSOE:W:CE SPECIFICATI0::S a

Calcu'_ated values of operating variables such as pressures, tempera-tures, flows, heat fluxes, rentivity coefficients and on-site radiation lcvels ur.dcr steady state and transient conditions which are stated in i

the sections of the license application listed.in Paracraph 3, above,

.d are ecnsidered to be performance specifications of the reactor and are H

incorporated by 2 eference herein.

Yankee shall not operate the f acility under circunstancos where there is a substantial variance between the d,

fo.eccing perforgance specifications and the corresponding values de-q

.ter.~.ir.ed by operation of the f acility.

M4 The performance and function of the systems described in the following E.)

secticns of the license application shall be substantially as descrioed;

'1 however, the detai.ls of individual components and their arrangement as described in each of these sections may be altered by Yankee at its own discretion provided that such an alteration would not violate some other provision of these Technical Specifications:

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-?t 206 Component Cooling Systen 208 Sampling System 211 Vent and Drain System, Primary Plant V p r Container Atmosphere Contr.ol Systens

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219 r' in and Ato:iliary Steam System a

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220 Condensate and Feed Water Syctem 221 Circulating Unter System' 222 Unter Supply System t

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.(9).The Com.ission shall be im.cdlately notified should an unexplained reactivity change greater than 0.0% d K/K take placc at any.timo sub:cquent to the first wock of i

full power operation.

This reprting recuirement shall be in effect only uhen.the' boron concentration in the primary system exceeds 60 ppm and within ono ucek aft.or a reduction to a boron concentration of 2c'ss than 60 ppm.

(10)

During a reactor startup in which core reactivity or con-i trol rod positions for criticality are n.ot established, a plot of inverse nultiplication rate (or count rato) versus rod position should be made, b.

Power Level (1) liith all four main coolant loops providing normal flow to the core, the reactor power level is limited to 600 Md thermal.

(2) With only three main coolant loops providin; normal flow to the core, the reactor power icvel is limited,to 1.50 RI thernal.

l (3)

The reactor uill be scrammed automctically by a lugh neutron fl=: 10. 01 cir.;n:1, ::t at n:t. ::r th:r.103;.', Of r t :i p:uct for caen condition se definta in (1) and (2) soove.

(h)

L: cept fcr operation of the reacter at niant, newer levels not exCceding l$ Ed cicctric, the reactor shall'not be oper-ated uith less than three unin coolant loops orovidin;: norr.a1 flow to the core.

(5)

Whenever there is a custained outage of one of the 115 kV lines because of naintenance g fault condition, the reactor power icvel chc11 be reduced to a 1cvel consistent with l

threc loop operation as defined in (2) cbove, c.

Thermal (1) During steady state power operation, the peak linear heat rate shall not exceed the limits shown in Figure 8-1.

With these liraits, if full power cannot be attained, the allow-l able f raction of full power shall bc calculated as followc:

Lirniting L11CR 12 Allowable Fraction of Full Fower = Peak Full Power LHGR, where the limiting L11CR is obtained from' Figure 8-1.

i The peak full power L11CR shall include the following:

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l n) Heat flux power peaking f actor, F measpred using core instrumentation at a power > 10%;

b)

Ef fect of inserting the control. bank from its pos'ition at the time of measurement to'its insertion limit (F ) as l

I shown in Figurc 8-2.

The rod insertion limit is shown in Figurc 8-3; c)

Ef f cet of xenon redistribution (1.10);

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d)

Flux peaking augmentation f actor (power spike) using l

Figure 8-4; t

e)

Shortened stack height factor (1.009);

f) Meas _rement uncertainty (1.05);

g)

Power icvel uncertainty (1.03);

I flux engineering factor, Fk, (1.04);

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Core average linear heat generation rate at full

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power (4.34 kw/ft).

s These f actors are. multiplicative and items (a) and (d) shall be c osen at a core height so as to' maximize their product.

When operating at const, ant power, all rods out, with equilibrium xenon, power peaking in the Yankee Rowc core decreases monotonically as a function of cycic burnup.

This has been verified by both calcu-lation and measurement on Yankee cores and is in accord with the expected beliavior in a core that does not contain burnabic poison.

The all-rods-out power peaking meas'ured at any time in core life thus provides an upper bound on ARO power peaking for the remainder of that cycle.

Therefore, the measured power peaking shall be checked every 1000 equivalent full power hours and the latest measured value shall be used in the computation.

The only effects whica can increase peaking beyond this value would be control rod insertion and xenon transients and these are accounted for in items (b) and (c).

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particulate monitor reading shall require bringing the plant to a hat standby condition to pe rrait primary systen inspection.

If both methods for determining main coolant Icakage L

become inoperative and an hourly :,chedule of local ccepling of vapor container air cannot he maintained, the plant will be broecht to a. hot standby condition.

c.

Ot,her Plsn, Prc ection (1)

The reactor chail be. r,crort.ed automat c a12y, above y-..,.

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electric, when the tu.-bir.c :.

e.r.pp:c it,r any hat or..

7ne turbine chail be prout ed b, sil uwal prow ::.*'O '. r t p S inc ludi ng 'rJ. Ch inr.;5 L befa r! nG k O~@ rL IbIC 10U cc~r:,0C oli s

pressure, lo,e ccndc..st r vacu r. Ar.c cVer:.ne20 -

.s (2)

The reactor t hall bc cror c:1 tu~ r ot' Cal.a 7 ' LD';'V C

' '".,n 2 ciect:tc wuen tne J.:r.eratcr :: L': P?e: l o r ~ : 7 l 0 0 7 7-gc f.c r a to r s r. ".ill O pr o t e t *. ' O *)

a '.1 V I L a 1,rer *etM t 1 VO t r '- p 3 inclucir.; overcut rent, c.ffcrernst or.: loss of f;eic.

( ~) ) Autou.at ic t al tiation of the sa f e th' inje c~t ion system, pumps and valves, shall be set to operate at a main coolant prescurc not

• lcss than 1700 psig or a containuent pressure not r. renter than 5 psin.

t (4) The Low Pressure Safety Injection Accumulator shall have a l

minimum usable water volume of 700 ft3 and a minimum nitrogen I

overpressure of 410 psig.

(5)

The Safety Injection System which includes the Charging Pumps i

shall be maintained in readiness to inject borated water into the reactor at all times when the main coolant pressure is 1,000 psig or higher.

Charging system readiness requires that l

two fixed speed pumps be availabic.

Since the pumps are high

.i, maintenance items, a fixed speed charging pump may be out of 4

I service for eight hours before a pressure reduction to less than 1000 psig is required.

+

i The following manually-controlled, electrically operated valves shall be in the position indicated during normal power operation.

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Valve No.

Normal Position PU-MOV-541 Open PU-MOV-542 Open PU-MOV-543 Closed PU-MOV-544 Closed PU-MOV-545 Open i

c PU-MOV-546 Open 4

PU-MOV-547 Closed PU-MOV-548 Closed "4

The following manually-controlled, cicctrically operated valves shall have power disconnected (both series breakers in open 3

position) to meet single failure requirements during normal power operation.

I Valve No.

Position j

1 CH-MOV-522 Closed l

CH-MOV-523 Open CH-MOV-524 Open (6)

The reactor shall be scrammed ' automatically, when the power level is above 15 MWe, by two or more low steam generator level signals.

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Control Pod Ecram - Circuit Chech - During plant operation it is not adviccble to taho any touts in the ceran circuitry. All i.ndications that are available will be noted at least once per shift, and adjuctnants will be nado only if necessary.

At no time will the scram circuitry be, disconnected during operation.

All rod scran tectc will be mcdo during ccheduled reactor shut-down or during various emergency chutdowns.

Safety Injection Syntera - Monthly, during power operation, all of the pumps and active valvec of the caf ety injection cyctcm will be, operated individually from the nafety injection control panel in the control room to deterraine their serviceability and correct light position indication.

The pumps vill 3

g be run to detcrninc.their starting capability only, because the loop fill and injection valves are open.

Whenever the reactor p3 ant is chutdown and deprennurized, the entiro tynten operation will be' checked by nanun3 operation _

-of the safety injection control cwitch provided on the nucicar nection of the a

main contrcl board.

Surveillance of the Safety Injection Actuation Signal (SIAS) Initiation Channeln will include:

(1) a comparison of the separate precourc indicationn by each shift; (2) a channc:1 calibration cach refueling by applying known precsuren to each sensor; (3) a system functional tent cach refueling by application of test signaln to each channel to verify system capability; t.nd (4) a monthly operational chocA of t.he tuo containment air prcccurc nwitchen.

'Prennerizer Enrav_Svrt om - In order to ansure reliab3 c perforrance, the prccourincr curge and circulatien sprey syntems vill be tested per.iodicajly preferably during the norr.al str.rtup oparation.

While maintaining normal cpori. ting conditions in the na~in coolant syctcm and the presnure contro) and relief t.ydte:a, that is with surge sprey donnergized, it in determincd that. cerrcct circulat. ion spray exists,(hand control valve ini tia31y ponitioned to ac to maintain preocuriner couiLibriun conditionc) by noting frequency of hontor cycling uhile maintainir.g proncuriner at normal valucc.

Verificat. ion of curge ryrt.y operation vill be acccmplini.ed by obtaining manuum spray flov and observing the resultant preasurizer pressure decrcose, terperature increase i

of water flowing through the surge line, and changen in proccuriner heater cycling.

Prcererinor rolenoid Relief Ya3ve - The pressuri: er t.olonoid relief valve,

.which in provided to limit the duty of the prensuriner safety valvec, vill be tcsted after refueling or after cenpletion of r.aintenance on this valve.

The set precouro and the bloudoun precourc shall be observed during the test operation and chall be compared with the design conditions.

After the valve hac discharged, note the downstrean pipe temperatures to' assure j

that the valve disc has properly receated.

The steam prescure required to test the solenoid relief valve ic obtained by operation of the pressurizer heaters, j

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