ML20127M830

Amend 3 to License DPR-22,including Change 14 to TS, Authorizing Partial Loading & Use of 8x8 Fuel

ML20127M830
Person / Time
Site:	Monticello
Issue date:	05/14/1974
From:	Goller K US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20127M816	List: ML20127M813 ML20127M830 ML20127M841 ML20127M857 ML20127M861
References
NUDOCS 9211300484
Download: ML20127M830 (26)

Text

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ATTAC11 MENT TO LICENSE AMENDMENT NO. 3 QiANGE NO.14 TO APPENDIX A TECH 41 CAL SPECIFICATIONS FACILITY OPERATDiC LICENSE NO. DPR-22 The following pages of Appendix A to Provisional Operating License have been revised to incorporate the changes related to safety valve modifi-cations, control rod scram times, standby gas treatment system, 8 x 8 reload fuel, and fuel densification.

Except as otherwise indicated, the enclosed revised pages supersede pages bearing the same nux:ser. The revised pages have marginal lines indicating where the changes appear.

Pages:

6 7

10 12 14 16 19 21 23 79 108A 108B - Replacing one unnumbered page issued S/24/73 (Change 9) 108C - Addition 112 113A N 113a / Replacing four of the unnurbered pages issued 8/24/73 (Ch. 9) 115 118

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NO3 TIES STNIES PGfdR CCePMN DOCKiT 1:0. 50-253 A'91"%T 'IO PRO'IISIC'AL O?EPATIMG LICETSr'

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/Wendment ::o. 3 Lic~ ue No. D?R-22 1.

The Ato-ic Energy Camission (the Cca,lssion) has found that:

A.

The applications for erenchent by the ':ortharn Shces Pover Comany (the licensee) dated I ov=rber 19, 1973, January 23, 3 974, and Ihrch 1,1974, as supplcranted, cogly with the standards and requira ents of the Atcmic Energy not of 1934,

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as amnded (the Act), and the Ccmission's rules and rcgulations sat fod in 10 C: Chapter I; B.

The f acility will onarata in ccnformibf with tha licanse, the provisions of the Act and the rules and regulations of the Cc rission; C.

'Ihre is reasona' ale assuranca (i) Sat the activitics authorized by this rerhent can be conducted without nndargering the health and safety of the public, and (ii) that such activities uill be conducted in complirce with the Ccanission's regulations; D.

The issuance of this cmndment will not be inimical to the cc=cn defence and secu-ity or to the health and safety of the public; E.

The request for a haring and petition for Ica;c to intervene (b-1 the Ilinnesota Pollutica Control Agency) on the prop 3s K1 action of those itcra relating to omration with 3 x S fuels and luitin: conditions for omration urecietM -ith fel densification for 2 a rd 7 :. 7 Z,als 'im; aer i.edrm;n j

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cr m se 10suas wi a tn.e licensing proceeding in /civing tn.a contersion of the provisicnal cparating license of the Mcncicello facility to a full tarm licanse (see:

Atcmic Safeb/ cnd Licensing Paard's t'ecorandum and Orc'ar %dirg on Petition for Lean to Intervene dated April 30, 1974), and F.

Prior twblic notice of those itcms relating to pressure relief, eg3_

trol rad scrar, tires, standby gas traatrent, and rea : tor vcscel te73ratu~e reasu' rents is not required since thyf do not irr/olve a significant haza-ds con 3Ldaration.

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Accordindy, parattraph '.i3 of Scility Licence.-io.

Di'?-22 t herehv v

'Tnded to read as follows:

"B.

Technical Saecifications to Technical Specificationo contained in Appendix A attached to Facility Operating License No. DPR-22 are revised as indicated in the attachment to this license i

acendment. h Technical Cpecifications, as reviced, are hereby incomoratel in tha license. 'Ihe license a chall operate the facility in accordance with the Tecnnical Specificatit..,.

2_ vised."

i 3.

Olia licence amuren; 13 effectiw aa of tne date of its issusnce.

4 iT]R E A'IUMIC EETt0Y OWLS 3IO:1

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Date of Issuance:

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Eases Continued:

2.1 The, design basis critical heat flux is based on an interrelatioaship of reactor coolant flow and steam quality.

Steam quality is determined by reactor power, pressure, and coolant inlet is a function of feedwater temperature and to a lesser degree reactor water invelenthalpy, which in

turn, tion is based upon experimental data taken over the pressure range of interest Thi:, correla-relation in a HWR, and the cor-line was very conservatively drawn below all the available data.

Since tha correlation line was drawn below the data, there is a very high probability that operation at the calculated safety limit would not result in a critical heat flux occurrence.

clad perforation would not necessarily be expected.

In addition, if a critical heat flux were to occur, mately 1100*F which is below the perforation temp'erature of the cladding material. Cladding temperature would i by tests in the General Electric Test Reactor (CETR) where This has been verified fuel similar in design t o Monticello operated above the critical heat flux for significant period of time (30 minutes) without clad perforation.

(1)

Curves are presented for two different pressures in Figure 2.1.1.

1he upper curve 1: based on a nominal operating pressure 1000 psig.

The louer curve I:, based on a pressure 1250 psig.

lu no case is reactor prersure ever expected to exceed 1250 psig, and therefore, the curves will cover all operating conditloas s'

with interpolation.

if reactor pressure should ever exceed 1250 psig during power operation asaumed that the safety limit has been violated.

it would be For pressures between 600 psin, which is the lowest prer sure us ?d in the critical heat flux data, and 1000 psig, the upper curve is :ppiicable with increaued margia.

The power shape assumed in the calculation of t hese curves was based on design liwi.tr and results in a l

total peaking factor of 3.03 for 7x7 fuct and 3.04 for 8x3 fuel.

For any peaking o f smaller magnitude, the curves are conservative.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuousl'y by the in-core Local Power Range Manitor (Ll'R11) System.

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To maintain applicabili ty of the safety limit curve, the safety limit will be loucred according to the equation given in Figure 2.1.1 in the rare event of power operation with a total peaking factor in execcc l

of the desi ;n value.

t (1)

T. Sorlie, et. al.

" Experiences with Operating IMR Fuel Rods above the Critical !! cat Flux" -

Nuciconics, Val. 23, :io. 4, April, 1965.

201 pASES 14 REV y

Eaoes Continued:

2.1 During transient operation, the heat flux vould lag behind the neutron flux due to the inherent heat I transfer time constant of the fuel.

Also,.the limiting safety system scram settings are at values which will not allow the reactor to be operated above the safety limit during normal operation or during other plant operating situations 1,hich have been analyzed in detail (4,5,6,7).

In addition, control rod scrams are such that for normal operating transients the neutron flux transient is terminated before a sigaificant increase in surface heat flux occurs.

Scram times of each control rod are checked. cach refueling outage to assure the inscrtion times are adequate.

Exceeding a neutron flux scram setting and a delay in the control rod action to reduce neutron flux to Icss than the scram setting within 0.95 seconds does not necessarily imply that fuel is damaged; however, for this specification a safety limit violation will be assumed anytime a neutron flux scram setting of the APRM's is exceeded for longer than 0.95 seconds.

Analysis within the nominal uncertainty range of all appropriate significant parameters, show that if the scram occurs such that the neutron flux dwell time above the limiting safety system setting is less than 0.95 seconds, the nafety limit will not be exceeded for normal turbine or generator trips, which are the nest severe normal operating transients expected.

The computer provided with Ibnticello has a sequence annunciation program which will indicate the sequence in which scrama occur such as neutron flux, pressure, etc.

This program also indicates when the scram set point is c1 cared.

This will provide information on how long a scram coadition exists and thus provide some

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measure of the energy added during a transient.

Thus, corsputer infornation normally wt 11 he availabic for analyzing scrans; however, if the computer information should not be avaliabic for any scram analysis, Specification 2.1.C.2 will be relied on to determine if a safety 1imit has been violated.

During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water Icvcl should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.

The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirda the core height.

Establichment of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will be continuously monitored whenever the recirculation pumps are not operating.

(4) FSAR Volume I, Section 111-2.2.3 (5) FSAR Volume III, Sections XIV-5 (6) Supplement on Transient Analyses submitted by USP to the AEC February 13, 1973 (7) Letter from USP to the AE", " Planned Reactor Operation from 2,000 !MD/T to cad of cycle 2", dated August 21, 1973 16 2.1 Willi EEV-

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au w toeh enoi t

nog e i ds ui a e a,m r 'm n hc int edr nr i onis as a3 c

i

,r aet u) f nr vct f.

c" f

• r u n

l i

r l

t mfe re e0 f' ms t

on i

u u

w ea onar t 8 nl ae h

o f c.

n x

rl r m

y c

w l.s e

g e

m p

i t

m

[

i ll r

Basen Contiuned:

2.3 the worat case MCllFR during steady state operation is at 1107. of rated pouer.

Penking factors as specified in Section 3.2 of the YSAR ucre couaidered.

The total penking factor was 3.05 for I

7x7 foci and 3.04 for 8x8 fuel.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPlui system.

As with the APRM scraa setting, the APRM rod block setting is adjusted downward if peaking factors greater than the design value exist.

This assures a rod block will occur before MCilFR becomes less than 1.0 cven for this degraded case.

The rod block setting is changed by increasing the APlui gain and thus reducing the slope and intercept point of the flow-biased rod block curve by the reciprocal of the AP101 gain change.

The operator will set the APB14 rod block trip settings no creater than that shon in Figure 2.3.1.

liowever, the actual set point can be as much as 3% greater than that shoun on Firure 2.3.1 for re-circulation driving flows less than 50% of design and 2% greater than that shown for recirculation driving flows greater than 50/o of design due to the deviations discussed on Page 18.

C.

Reactor Iow Water Level Scram - The reactor low water level scram ia net it. a point which will i

assure t, hat, the vtt,cr 1cve] used in the bance for the safet.y 'li;ait is maintainet.

The operator will set the low water level trip setting no ]ower than 10'6" aL the top of the

.c active fuel.

However, the actual set point can be as much as 6 inches lower due to the deviations discussed on Page 18 D.

Reactor Low Low Water Level ECCS Initiation Trip Point - The (mergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the cncrgy ar.nociated with the loss of coolnut accident and to limit fuel clati temperat.ure to well below the elat melting temperature to assure that core ccometry rer.ains intact and to limit any clad metal-water reaction to less than II. The design 'or the ECCG comfonents to meet t.he above criterion was depen ient, on three previously set parameters:

the maximum break size, the low water level sc ran act point, and the ECCS initiation set point. To lower t he set point for ini ti ation of the T:CCS could prevent the ECCS ccmponents from meeting their criterion. To raise the ECCS ini ti ation set, point, would be in a safe direct. ion, but it would reduce the mugin established to prevent actuation of the fiCC3 during normal operation or during nonaally expected transiruts.

-2. 3 P.Af,12; P1

y

l i

i 1

i i

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=

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2.0 SAFETY LIMITO LI!'1 TING SAFETY SYSTat LETTINGS l

r i

i

,l i

202 REACTOR 0001AUT SYSTEtt 2.4 FiEAcrOR 000IAITP SYSTEM.

)

l L

Anplic*tbili ty:

I A glicability:

l j

Applies to limits on reactor coolant system t

Applies to trip settings of the instrument i

pressure.

I and devices which are provided to prevent the reactor system safety limits from being. ex-l ceeded.

(

]

ObPetive:

Objective:

i h

)

To establish a limit below which the integrity

?

j of the reacter coolant system is not threatened To define the level of the p.ocess variables.

due to an overprescure condition.

at which auto:mtic protective action is initiated to prevent the safety limits from being excaeded.

I i

Srecifi cation:

Sreci fica ti on:

l The reitetor vessel pressure shall not exceed A.

Reactor Coolant High Pressure Scram shall i-1335 paia at any time wLen irradiated fuel is be 3-1075 psig.

{

present in the reactor vassel

)

B.

Reactor Coolant System Safety /Relier Valves ~

shall be set as fo11cus:

+

l 8 valves at i 1080 psig.

i ii l

+

1 i

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l 2.2/2.4 l

2.3 l

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3.0 LI!!ITING CCNDITIONS FOR OPEPATION 4.0 SURVEILIANCE REOUIREMENTS Io Recirculation System I.

Recirculation Systcm 1.

Except as specified in 3.5.1.2 below, wheaever

1.. Once per tonth, uhen irristed fuel is in tl...

Irradiated fuel is in the reactor, with reactor reactor with reactor coolant temperature grent' coolant temperature greater than 212 P and both than 212 F and both reactor recirculation resctor recirculation pumps operating, the recirculation system cross tie valve interlocks pumps operating, the recirculation' system crom tic valve interlocks shall be demonstrated to shall be operable.

be operable by verifying that the cros tie valves cannot be opened using tInw nmT.zal coat:r 2.

The recirculation system cross tie valve inter-switch.

j locks may be inoperable if at least one cross

~

tie valve is maintained fully closed.

2.

When.a recirculation system cross. tie valve l

Interlock is inoperable, the position of at I

least one fully closed cross tie valve shall I.

Averaga Planar L11CR be recorded daily.

Dur 6g steady state power operation, the average iincar heat generation rate (LHCR) of all the J.

Average Planar LIICR rods in any fuel assembly, as a function of average planar exposure, at any axial location.

Daily during power operation, the average shall not exceed the maxicua average planar DICR planar LIICR shall be checked.

showr. In Figure 3.5.1.

3.5/4.5 103A REV 1

2 1

I I

h

. 5 j

3.0 LIMITING CONDITIONS I'OR OPERATION 4.0 StiRVEILIAt;CE RE(IllREENTS i

i e

h F

K.

Iacal UlGR K.

Incal UIGR During steady state power operation, the linear Daily during reactor power operation, the j

hest generation rate (U1CR) of any rod in any local U!GR shall be checked.

fuel assea.bly at any axial location shall not exceed the maximum allowable UICR as calculated by the following equation.

I i

UICR 5

UlGR 1IAP I L i

max d

P max

,LTj i

3 DICR r

j do Design UICR l

k 17.5 ku/ft for 7x7 fuel

=

13.4 ku/ft for 3x8 fuel

=

I

' AP max = Ihxinnun power spiking penalty i Pl 0.033 for 7x7 fuel 0.024 for 8x8 fuel

=

LT a Total core Icagth =>l2 ft L - Axial posit'on above bottom core 1

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i Ilnsen Continued:

P l

rargin, the RCIC. system (a non-safeguard system) has been required to be operable during this ti::0, since the RCIC system is capable of supplying significant water makeup to the reactor (h00 gpm).

i i

E.

Automatic Pressure Relief j

The relief valves of the autcmtic precrure relief subsystc.a are a backup to the IITCI subsystem.

[

They enable the core spray nystem or I.PCI to provide protection against tha scall pipe break in the event of IIPCI failure,,by deprcccurising the reactor veasel rapidly caough to actua te the care j

sprays or IJCI.

Either of the two core spray syst:s.s or ITCI provide suf ficient flow of coolant to l

limit fuel clad temperatures to well below clad nelt and to assure that core geometry recains intact.

l

.Three s'af ecy/ relief valves are included in the automatic pressure relief system.

Of these three, only two are requir(d to provide suffi' *, cut capacity for the automatic pressure reliefsystem.

Sec l

section 4.4 and 6.2.5.3 FSAR.

.7.

PCIC I

The FCIC system is l rovided to sorply continuous makeup water to the reactor core when the reactor i

is isolated frcs. the turbine and when the feedwater cystua is not available. The pumping

(

capacity of the RCIC system is sufficient to mintain the v ster level noove the core without any other Vater syste:1 in operation.

If the water level in the reactor vessel decreaccc to the RCIC initiation level, the system autom:.tically starts. The system may also be manually initiated at any time.

Tbc 1[PCI system provides an alternate nethed.of supplying makeup water to the reactor should --

l the normal feedvater become unavailable.

Therefore, the specification calls for an opera-tility check'of the liPCI cystem should the I!CIC cystem be founi to be inoperable.

I 5

i 1

4 3.5 faSES 112 RSV 1

'l

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

,m.

t

?

3 k

Itnnes Continued 3.h J.

Average Planar LIICR i

i, 4

This Specifica: f on assures that the peak cladding te=perature following the postulated design basis

)

loss-of-coolant accident will not. cxceed the 2300"F limit specified in the Interim Acceptance Criteria i

)

(IAC) issued in June 1971 considering the postulated effects of fuel pellet densification.

l.

The peak cladding temperaturc follouing a postulated loss-of-ccolant accident is primarily a function j

of the average heat generation rate of all the rods of a fuel assembly at aay axial location and is l

only depcudent secondarily on the rod to rod pouer distribution within an assenbly.

Since expected local variationa in power distribution within a fuel assembly affcet the calcu lated peak clad temper-ature by less than i 20 F relative to the peak tecperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures ara belou the IAC limit.

1 The maxiuum average planar LHCR curves shown in Figure 3.5.1 were calculated for the :various Ibnticello fuel types in the manner discussed in Section 4.3 of General Electric topical report, "GECAP-III:

A

?bdel for the Prediction of Pellet-Cladding Thcrmal Cbnductance in EUR Fuel Rods", NEDO-20lS1, Revi sio n 1, I;ovember 1973.

These curves show the composite limitation based on the design LHCR of the. fuel and the peak cladding temperature in the event o f a LDCA.

Calculations based on the AEC " Modified CE Ibdel for i

Fuel Densification" attached to a Dece:bar 5, 1973 letter from D J $kavhols (iG AEC) to L 0 !!ayer (NSP).

4 1

The possible ef f ects of fuel pellet densification were:

(1) creep collapse of the caldding due to i

axial gap lonnation; (2) increase in the LilGR !>ccause of pellet column shortening; (3)pover spikes

[

1 due to axial gap fornation; and (4) changes in stored energy due to increased radial' gap size.

Cal-j j

culations show that clad collapse is conservatively predicted not to occur currently or pr.ior. to l

Septen.ber 1974.

Th e re fo re, clad collapse is uot considered in t he so alyse:.

Since axial t h erma l-expansion of the fuel pellets is greater than axial shrinkage due to densificat. ion, the analyses oi peak clad temperature do not consider any change in Ll!GR due to pellet column shortening. Although, l

the formation of axial gaps might produce a local power spike at one location on any one rod in a -

\\

fuel assembly, the increase in local powcr density would be less than 27. at the axial midplane.

Sinc e i

small local variations in power distribution have a small effect on peak cl'ad reaperature, power spikes j

[

were not considered in the aaalysis of loss-of-coolant accidents.

Changes in radial gap size affect 1

i i

i i

.i 3.5 nASES 113^'

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P"U

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=-

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3 0 LI111 TING CO!iDITICIIS FOR CPI 3tATICU h.0 LURVEILIJJ'CE REQUIRE !.TS i

306 PRUmRY SYTcDM ICU: miry 4.6 PRDWC SYSTS! IGHWARY Applicabil i1 y:

Applicability:

Applies to the operating status of the reactor Applies to the periodic examination and testing coolant systes.

requirecents for the reactor coolant system.

Objective:

Oldectivc :

To assure the intcgrity and safc operation'of the To deternine the ccndition of the reactor coolant reactor coolant cystem.

system and the operaticn of the safety devices related to it.

e Specificaticn:

Cpecificaticn:

A.

Therr11 Limitations A.

Themal Limitations 1.

Tha averace rate of reactor coolant 1.

Daring heatup:, and cocidovna recirculation temperature change during norral heatup loops A and B temperatures shall be per-or cooldc*.m clall not exceed 100 P/hr.

manently recorded at 15 minutes intervals.

when averace:1 over a enc-hour period.

2.

The punp in an idle recirculation locp 2.

The temperatires listed in 4.6.A.1 shall shall not te started unlecs the te=per-be permanently recorded subsequent to a ature of the coolant within the idle re-heatup or coolcown at 15 minute intervals 0

circulaticn lecp is within 50 F of the until three consecutive readings are withir reactor coolcnt tenperature.

5 degrecs of each other.

~ 3 6/4.6 115 rev

=----.w...-v-a,... -

N

]

1 l

r i

a i

i 30 LIMITING CCHDITIOUS FOR OPEi?ATION 4.0 51m_VE_I.Il1J!CE RERUIIiEFl:TS

[

j l

(b) When the ecntinuous conductivity coni-tor is inoperable, a reactor ecclant sample shculd be taken at least once.

1 i

per shift and analyzed for ccnductiv-

,i ity and chloride ion centent.

I I

4.

If Specificatica 3.6.c.1, 3.6.C.2, and 3.6.

~

C.3 are not cet, normal orderly shutdown i

4 i

uhall te initiated.

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1 D.

Coolant Lcakage D.

Coolant Leakage

[

i 4

Any time irradiated fuel is in t.he reactor vessel, Reactor coolant system leakage into the dry-

.i l

and reactor coolant temperature in above 2120F, ucll shall be checked and recorded at least renctor coolant lcakage int o the primary ennt ain-once per day.

4 ent from unidentified nourcen chall not, exceed i

I 5 t'.lun.

In addition, the tctal reactor coolant I

j system leakage into the primary containment shall i

not exceed 25 m m.

If these conditions cannst te met, initiate an orderly shutdown and have the re-i actor placed in the cold shutdown condition within i

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

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3.6/4.6 113 i

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30 LIMITIIiG 00!IDITIO!IS FG2 OPEP.ATIO:I 4.0 SURVEII.Ild:CE P1 1'JIREMEITIS 1

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F.

Structural Intecrity F.

Structural Integtit-/

i I

The structural integrity of the primary system

'IM nondestruct * ' inspections listed in Table l

boundary shall bc maintained at the level re-4.6.1 -Jall te );erronned as specified. 'Ihe j

quired by tha original acceptance standards results oMained from cocpliance with this throughout the life of the plant..

specificatio0 vill te evaluated after 5 years

[

and the ccnclusionn of this evaluation vill l

te revicwed with the AEC.

r j

G.

Jet Pumps G.

Jet Picps l

]

• Whenever the reactor is in the Startup Whenever there is recirculation flev vith the or Run modes, all jet purps shall be oper-reactor,in tie Startup or Run modes, jet pucp j

able. If it is determined that a jet pump is operability chall 1 e checked daily by verify-inoperable, the plant shall be placed in a ing that all the following conditions do not cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

occur simultanecusly:

I 1.

Tne two recirculation loop flows are untalanced by 15),or more when the recirculation pumps are operating at the same speed.

a l

2.

The indicated value of core flow rate is 10% cr more less than the value de-a rived frc;a loop flow censurements.

j

3. 6/4.6 120 16/

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t Bases Continued 3.6 and 4.6:

i

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i D.

Coolant Leakarc The former 15 gpm limit for Icahs frca unidentified sources was established assuming.such leakage was coming l

from the primary system.

Tests have been conducted shich demonstrate that a relationship exists between the ' size of a crack and the probability that the crack will propagate.

Fron the crack sisc a leakage rate can be determined.

For a crack size which gives a leakage of 5 gpm, the probability of rapid propagation is less than 10-5 Thus, an f

unidentified leak cf 5 gpm when assumed to be f rom the primary system had less than one chance. in 100,000 of propa-

[

gating, which provides adequate nargin. A leakage of 5 gpm is detectable and ceasurable.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period I

allowed for determination of Icakage is also based on the lou probability of the crack propagating.

l I

i i

The capacity'of the drywell su=p pumps is 100 gpa and the capacity of the drywellL cquipment drain _ tank pumps j

is also 100 gpm.

Removal of 25 gpm frca either of these su=ps can be accomplished with considerable,=argin.

4 i

An annual report will be prepared and submitted to the AEC summarizing the primary ccolant to drywell leaka3c j

measurements.

C:ter tect.niques for detceting icaks and the applicability of these techniques te Sc Monticello l

Plant will be the subject of continued study.

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j E.

Safety / Relief Valvec g

]

Testing of all safety / relief valves each refueling outage ensures that any valve 'dctcrioration'

}

j is detected. A tolerance value of 1% for safety / relief valve setpoints is specified in Section III of the

. AS!E Boiler acc Pressure Vessel Code. Analyses have been performad with all valves assumed set 1% higher j

(1030 psig + 17.) than the nomlnal setpoint; the 1375 psig code limit is not exceeded in any case.

l l

The safecy/ relief valves ave used to limit reactor vessel overpressure and fuel thernsi duty.

\\

The required safety / relief valve steam flow capacity is determined by analyaing the transient t.ccompcaying the mainsteam flou stoppnge resulting from a postulated IGIV Closure from a pawer of 1670 12.

The analysis t

i assumes a multiple-failure wherein direct scram (valve position) is neglected. Scram is, assumed to be from indirect means (high flur).

In this event,: the safety / relief valve capacity is assumed to bc 71*4.of the full j

power steam generation rate.

[

}

3.6/4.6 LASES 134 t

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' 3.0 LIjGTII;G CO!iDITIOi!S FOR OFERATICII 4.0 SURVEILIld:CE REQUIRElE!iIS i

6.

If the specifications of 3.7. A cannot pc met, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

N B.

Standby Gas Treatment System i

B.

Standby Cas Treatment System 1.

Except as specified in 3.7.B.3 below, both circuits of the standby gas treat-y g s treatment sysgem surveillance an ment systen shall be operable at all shall be performed as indicated below:

times when secondary containment integrity is rcquired.

a.

At 1 cast once per operating cycle it 4

shall be demonstrated that:

4 (1) Pressure drop across the combined i

high-efficiency and charcoal filters is less than 7.0 inches of water, and (2) Inlet heater output is at Icast 15 kw.

b.

During each refueling outage prior to refueling, whenever a filter is changed, i

3 whenever work is performed that could I

affect filter systems efficiency, and at I

intervals not to exceed six months between refueling outancs, it shall be demonstrated f

that:

i.

^

(1) The removal efficiency of the installed r

l particulate filters for particles having l

4 a maan dicaater of 0.7 nicrons shall be.

3.7/4.7 -

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I 30 LIMITING CONDITIONS FOR OPER/sTION h.O SURVEILL" dice RIQUIRE!Df13 3

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1 equal to or greater than 997. based on

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m an in-place dioctyl phthalate (EOP) test.

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(2) The removal efficiency of the

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charcoal filters is not less than 997 for freun based on a freon test.

a i

c.

At least once each five years removabla l

charcoal cartridges chall'be removed and

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l adnorption shall be demonstrated.

l d.

At least cnce' per oper atin6 cycle automatic

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initiation of: each branch af the standby gas treatment system shall be demons trated.

2 2.

From and af ter the date that cne circuit 2.

L' hen one cir. uit of the standby cas treatment

}

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of the stein.lby c tc treatment cyctem in yctem becomcc inoperable, the operable j

l, made or found to be inopenable for any circuit includinc its emergency power'saurce reason, re'ictor operation is permincible chall be denonstrate.1 to be operable icacdi-only during the succeedinc seven days un-ately. The operable circuit of the Standby

[

less such circuit is sooner ande operable, Gas Treatment System chall' be demonstrated 3

provided that during cuch seven days all to be operable daily thereafter.

.)

)

active ccerponentn of the othrer standby gan treatment circuit including its emergency power caurce chail be openable.

L r

i 4

3 If thic condition cannot be met, procedures shall te initiated immediately to establich the conditionc listed in 3 7.c.l. (a) throuch (d), und compliance chall be ccmpleted l

within 24 haurs thereafter.

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