ML20085C999

From kanterella
Jump to navigation Jump to search
Proposed Tech Spec 4.19 Re Auxiliary Bldg Crane Lifting Devices
ML20085C999
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 10/04/1991
From:
NORTHERN STATES POWER CO.
To:
Shared Package
ML20085C994 List:
References
NUDOCS 9110150181
Download: ML20085C999 (30)
v • d • e


Text

l Exhibit B Prairie Island Nuclear Generating Plant License lunendment Request Dated October 4,1991 Proposed Changes Marked Up On Existing Technical Specification Pages Exhibit-B consists of existing and new Technical Specification pages with the proposed changes highlighted on those pages. The existing pages affected by this License Amendment Request are listed below:

TS-lit TS . 'ri TS-vii TS-xi TS.3.8-3 Table TS.4.1 2B (page 1 of 2)

Table TS.4.1 2B (page 2 of 2)

TS.4.19 1 TS.S.6 1 TS.5.6-2 TS.S.6 3 B.3.8 1 B.3.8 2 B.3.8-3 B.4.19-1 l.

l.

l l

l.

l l

l l- 9110150181 911004 PDR l

P ADOCK 05000282 i

pygg 1

TS 111 TABLE OF CONTENTS (Contirried)

IS SECTION 11I11 PACE 3.6 Containment System TS.3,6 1 A. Containment Integrity TS.3.6 1 B. Vacutun Breaker System TS.3.6 1 C. Contairnent Isolation Valves TS.3.6-1 D. Conteinment Purge System TS.3.6 2 E. Auxiliary Building Special Ventilation Zone Integrity TS.3.6-2 F. Auxiliary Building Special Ventilation System TS.3.6-3 C. Shield Building Integrity TS.3.6-3

11. Shield Building Ventilation System TS.3.6-3 I . Contaitunent Internal Pressure TS.3.6-3 J. Containment and Shield Building Air Temperature TS.3.6-4 K. Containment Shell Temperature TS.3.6-4 L. Electric ilydrogen Recombiners TS 3.6-4 M. Cont ainment Air Locks TS.3.6-4 3,7 Auxiliary Electrical System TS.3.7 1 3.8 Refueling and Fuel llandling TS.3.8-1 A. Cora Alterations TS.3.8-1
5. Fuel llandling mal-Grene Operations TS.3.8-3 C. Small Spent Fuel Pool Restrictions TS.3.8-4 D. Spent Fuel Pool Special Ventilation System TS.3.8-4 E. Storage of Low Burnup Fuel TS.3.8-4 3.9 Radioactive Effluents TS.3.9-1 A. Liquid Effluents TS.3.9-1
1. Concentration TS.3.9-1
2. Dose TS.3.9-1
3. Liquid Radwaste System TS.3.9-2
4. Liquid Storage Tanks TS.3.9-2 B. Caseous Effluents TS.3.9-3
1. Dose Rate TS.3.9 3
2. Dose from Noble Cases TS.3.9 3
3. Dose from 1-131, Tritium and Radioactive Particulate TS.3.9-4 4, Caseous Radwaste Treatment System and Ventilation Exhaust Treatment Systems TS.3.9-4
5. Containment Purging TS.3.9-5 C. Solid Radioactive Waste TS.3.9 6 D. Dcsc from All Uranium Fuel Cycle Sources TS.3.9-6 E. Radioactive Liquid Effluent Monitoring Instrumentation TS.3.9-7 F. Radioactive Gaseous Effluent Monitoring Instrumentation TS.3.9-7

. c.

TS vi-

_IABLE OF CONTENT 5 (Contit.oed)

I

~TS SECTION' TITLE PAGE 4.12 Steam Generator Tube- Surveillance TS.4.12 1

-A,-Steam Generator Sample Selection and~ TS.4.12-1

-Inspection 4 B. Steam Cencrator Tube Sample Selection TS.4.12 1 I and Inspection.

C. Inspection Frequencies TS.4.12 3 D.-Acceptance Criteria TS.4.12 4 E.' Reports- TS.4.12-5 4.13 Snubbers TS.4,13-1

.4.14 Control Room Air Treatment System Terts TS.4.14 4.15 Spent Fuel Pool Special Ventilation System TS.4.15 1-4.16 Fire Detection;and Protection Systems TS.4.16-1 ,

A. Fire Detection Instrumentation TS.4.16 1 l

!. B. Fire Suppression Water System TS.4.16el. l l C. Spray and Sprinkler Systems' TS.4. i l D. Carbon Dioxide System TS.4 16 31 l E. Fire Hose Stations TS.4.16-3 1 F. - Fire Hydrant ilose llouses TS.4.16-4 ,

C. Penetration Fire Barriers TS,4.16-4 4.17 Radioactive Effluents Surveillance TS.4.17-1 A. Liquid Effluents TS.4,17-1 B.-Gaseous Effluents TS.4.17 2 C. Solid Radioactive Waste TS.4.17 4 D, Dose from All Uranium Fuel Cycle Sources TS.4.17-4

'4.18L Reactor Coolant Vent System Pathc TS.4.18-1 l; A. Vent. Path Operability TS.4.18 1 l

B. System Flow Testing TS,4.18-1 4/193[Ankilibry[B.uilding[_Crsn% Lifting l Devices? ' < jTS.4jl9T1 l '.

l' i

[.

t.

u

TS-vii TABLE OF CQb'IENTS (Cnntinued)

TS SECTION IIILE PACE 5.0 DESIGN FEATURES TS.S.1 1 5.1 Site TS.5.1 1

-5.2 A. Containment Structures TS.5.2 1

1. Containment Vessel TS.S.2-1
2. Shield Building TS.S.2-2
3. Auxiliary Building Special Ventilation Zone B. Sr'cial Ventilation Sy tems TS,5.2 2 C. C iinment System Functional Design TS.S.2-3 5.3 Rear TS.S.3 1

~ ce 18.5 3-1

. - ant System TS,5.3-1 C. . veteme- TS.S.3-1 5.4 Engins- ,

-tures TS.S.4 1 5.5 Radio '

-- ss TS . S . 5 1 A . ns-; TS.S.5-1 B. Rout. TS.S.5 1

1. Liq. . . . . . TS.S.5-1
2. Cascous Vasem. TS.S.5 2
3. Solid Wastes TS.S.5 3 C. Process and Effluent Radiological Monitoring TS.5.5-3 System 5.6 Fuel llandling TS.S.6 1 A. Criticality Consideration TS.S.6 1 B. Spent Fuel Storage Structure TS.S.6 1 C. Fuel Handling TS.5.6-2 D. Spent Fuel Storage Capacity TS.S.6 32

t-

-r ( ' t .-

-TS ui TABLE-OF CONTENTS (continuedl

-T,$ BASES SEQ 11QH TITLE ,_1101 ,

4 [0 BASES FOR SURVEILIANCE REQUIREKENTS 4.1 Operational-Safety Review B.4.1 1  ;

4 ~. 2 - Intervice Inspection and Testing of Pumps B.4.2 ,

and Valves Requirements -J 4.3 Primary Coolant System Pressure Isolation B.4.3 1 '

Valves l 4.4 Containment System Tests B.4.4 1 4.5 Engineered Safety Features B.4.5 ,

4.6 Periodic Testing of Emergency Power Systems B.4.6 1 l 4.7' Main Steam Isolation Valves B.4.7-1. l 4.8' -Steam and-Poser. Conversion Systems B.4.8-1 i 4.9 Reactivity Anomalies B.4.9 1 l

.4.10 Radiation Environmental Honitorin6 Program B.4.10-1 '

A. Sample-Collection and Analysis B.4.10 1 L B. Land Use Census B.4.10 1 l C. Interlaboratory Comparison Program B.4.10 1 4.11' Radioactive Source Leakage Test B.4.11-1 4.12 Steam Generator Tube Surveillance B.4.12 1 4.13 ' Snubbers B.4.13 4.14 Control Room Air Treatment System Testa B.4.14-1 4.15 Spent Fuel Pool Special Ventflation System B.4.15 1

.4.16 -Fir? Detection and Protection Systems B.4.16-1 4.17 Radioactive Effluents Surveill.mce .B.4.17-1 4.18 Reactor Coolant Vent System Paths B.4,18-1 BI19 M6x111 a ry]; Bu il dinf C r rineJ I.i_f tilngTDevic e s? sBTM1931 L

l l

l

,,: nu ar

--TS.3.8 3 3.8.B.. Fuel Handline ^nd "r:r- Doeratiofg

1. - During" fuel handling- operations w-+wme aperet 12: u!4h 1 :d:

.:ver spent fuel (!neid: th: epent feel peel: enclecure), the

- following conditions : shall be ' satisfied:

y _ . . . .- .

a. ' Radiation levels in the-spent fuel storage pool area shall

,be monitored continuously during fuel handling operations.

h, "rier : thereducing r sp ,t fuel chipp4*g-eask4*4+-t4.e pent fuel peel err + -

(1) ^ :!-ninam -bwer. car e nt+at4 r 'ef 1800 ppr ch:11 he metateined t- Opent-fuel prel: ":, 1 .nd ', Phe-regelred b:r- concentretier -hell be :rified by iehem44:n-enalye4+-da44y-wh44c' ca: o f - :4u cok' een44eoe*r-end

( 2 ) '. cach-4 pse t=14*14er de t e r:4ned-te he ceprble-+f i

. absorbing th i=pcet- energy f---a ::N f r p er r ersch p:d : pah1: _ ;f c';;e r.b4 ng the tapaet--eeert,y-et-e-eeek .

dr p ch:11 5: 1: ple::, and (2) cre- - irter4" '#el step: I ki-t i ng t ravet .

th: 2 d lead path e' '^ "-tw a4med te 5:

n.u. t n m..,_.t

. - . .a.7p r ev e

44) Fue! !n *he -::all pud--(p; ,1 " 1) ch:11 './ been dicchrg:.d fr- e receter f+r " least 5 yea rs,-

.b. Prior _to fuel handling operations; fuel-handling cranes shall-be load-tested for OPERABILITY of limit switches, ir.terlocks and alarms.

4. "han-th: spr,t fuel _cach centaine er er ar: fuel
b1!: , it .3111 not he sucpr'ded mere than 'O feet

^'- ny-swfeee-unt44-4hc furl ' 0 decayed :nre than oO day *

io_nrot,: 1800;. ppa._.,

s.s. hallflie.

e 4.

,%.wn ,mi.i_m.umTh.-

nn y a.

oronTeuncentrat

n. ..

. ac e- ..

,c.c- .

main. tained;in,T the...o' spent 3fu:e_lcpoob,wh,,ne

t. e + c.as -

co_n_ta.ih. in5_;fue_lns.e,l.oc. e t. s,,~t.ed P_e ntt f_u.kin _5th,w..~. ,v.er; a;spenn,

~ .

.c .

. - _4 a - - e_ll. P_oo.~l v,

2. If'any of the conditions.in 3.8.B.1, above, cannot be met, suspend sts i-all fuel handling operations Endlihlt.1;se;se?thbTaEtions3ss6shiaMtsi Ed; hs..Ea. bT..-~ ikh?E6.. f.spli- aisca.~M. ith~ hh.e. he. A' u. ~i ~ _nts~N fB.=3,Y8.iBil and arene

-~ ~ - -

= u .s 4 ira s s a g ,. , w. gv g w  %"e. -

ene4+ ewe}.

e Table TS.4.1-2B (Page 1 of 2)

TABLE TS.4.1-2B MINIMUM FED 11RJfRS FOR SAMPLING TESTS FSAR Section TEST FREQUENCY Referenef_

l. RCS Gross 5/ week Activity Determination
2. RCS lsotopic Analysis for DOSE 1/14 days (when at power)

EQUIVALENT 1 131 Concentration

3. RCS Radiochemistry E determination 1/6 months (l) (when at power) 4 RCS Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever Including 1-131, I-133, and 1-135 the spacific activity ex-ceeds 1.0 uC1/ gram DOSE EQUlVALENT 1-131 or 100/E uC1/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> fol' lowing THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown)
5. RCS Radiochemistry (2) Monthly
6. RCS Tritium Activity Weekly
7. RCS Chemistry (Cl* , F* , 02 ) 5/Veek
8. RCS Boron Concentration *(3) 2/Vock (4) 9.2
9. RWST Boron Concentration Weekly
10. Boric Acid Tanks Boron Concentration 2/Veek 11 Caustic Standpipe Na011 Concentration Monthly 6.4
12. Accumulator Boron Concentration Monthly 6
13. Spent Fuel Pit Boron Concentration Monthly (7) 9,5.5

Table TS.4.1 2R (Page 2 of 2)

TAB 1.E TS.4.1 2B MINIMUM FREQUENCIES FOR SAMPLittG_ TESTS FSAR Section JEST FREQUENCY Re fe rens.i, _

14. Secondary Coolant cross Weekly Beta-Gamma activity
15. Secondary Coolant Isotopic 1/6 months (5)

Analysis for DOSE EQUIVALENT I 131 concentration

16. Secondary Coolant Chemistry Ph 5/veek (6)

Ph Control Additive 5/ week (6)

Sodium 5/ week (6) ,,,

Notes:

1. Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
2. To determino activity of corrosion products having a half-life greater than 30 minutes.
3. During REF"JELING, the boron concentration shall be verified by chemical -

analysin daily.

4. The maximum interval between analyses shall not exceed 5 days.
5. If activity of the samples is greater than 10% of the limit in Specification 3.4.D, the frequency shall be once per month.
6. The maximwn interval between analyses shall not exceed 3 days.
7. ;Theidinimum? spent fuel pool boron concentration from Specification 3.S;B.1.b shallibe verified by chemical analynia weekly while a spent fuel _

cask containing.fue1Lis located in the spent _ fuel pool.

  • See Specification 4.1.D

TS.4.19 1 4.19 Auxiliary Butidinc Crane Liftine Devices 6Dtdicability Applies to surveillance requirements for the auxiliary building crane special lifting devices and slings before handling heavy loads carried over safe shut:down equipment.or spent fuel in the spent fuel pool.

Objective To ' verify ' that special lif ting' devices and slings' used in conjunction with ' the auxiliary building. crane are operable prior to their use in supportingLheavy loads: over safe shutdown equipment or ' spent fuel in;the spent fuel pool'.

Suecification Slings and"special lifting devices vuich will be used in supporting heavy loads from the auxiliary building crane shall be visually inspected and verified OPERABLE within 7 days prior to their use in handling heavy loads over safe shutdown equipment or spent fuel in the spent fuel pool.

4 TS.5,6 1 5.6 FUEL HANDLING A. -Criticality Consideration The-new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class 1 (saismic) structures. The spent fuel pit has a stainless steel liner to ensure against lons of water (Reference 1).

The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locationn.

The design of the new fuel storage pit and racks (Reference 1) ensures a new fuel pit K,n oi less than or equal to 0.95, including uncertainties, even if unborated water were used to fill the pit. The new fuel rack

. configuration also ensures K,g less than or equal to 0.98, including uncertainties, even if the new fuel racks were accidentally filled with a low density moderator which resulto' in optimum low density moderation conditions. Fuel stored in the new fuel storage racks wil.1 have a maximum enrichment of 4.25 weight percent U 235.

The spent fuel storage rack design (Reference 1) and the limitations on the storage of low burnup fuel contained in Technical Specification Section 3,8.E ensure a spent fuel pool K,g of less than or equal to 0.95, including uncertainties. The maximum enrichment of iuol to bu stored in the spent fuel pool will be 4.25 weight percent U-235.

T4*e-er4+ 4ea14+y-eena44+++ t4 me-oe- t4 wy-re4* t *r dawirepid ng-o f--*

epent4w4-ec ek ( ! r . , heavy-lead-)-drep-ont o-t4m-reeko4.ac -- bee: ava4*-

t4,mi-Oy400-fw atwi. TN max 4mu%74 * - beca-+alenletwi-ta.-4:r h water /"Og ret 12: f : hetwo m ? ^ and. ? . 2 u ! 44 -*4 e++n..concent+*t4ms4 18001 pper fur 1?i1EhBtGbe?frisisitsdiintuj a 'spenCfder cask 71rcthe poolSantessf a F iniibjumi heronj coscest r ationl o fi 1800 v ppm ; is1 p re s ent . 3The-1800 ppm Will eLnsure (tha ti kg forf the ' spent ' fuelicask, (including is csti Stical

' ncertaintisagwil1%elless u thanfor^egaalito 0195 for';a11Epostula~ted o

arrangementsLofl fuel withine thn cask.1The criticality? analysis forJ the TN401sifeht(fuelfstorbelea.sk;wasbased;on)frech-fuelenrichedfto;3i85 poighttpe m nt & 235)

B. Snent Fuel Storne Structure

! The spent fuel storage pool is enclosed with a reinforced concrete building having 12- to 18 inch thick walls and roof (Reference 1).

The r. d and pool enclosure are Class I (seismic) structures that afiw protection against loss of integrity from postulated tornado missiles. 1he storage compartments and the fuel transfer canal are connected by fuel transfer slots that can be closed of f with pneumatically sealed gates. The bottoms of the slots are above the tops of the active fuel in the fuel assemblies which will be stored vertically in specially constructed racks.

j. The spent fuel pool has a reinforced concrete bottom slab nearly 6 l feet thick and has been designed to minimize loss of water due to a

( dropped cask accident. 4n-midtt4+nr -t4w-upeut-fw4-cook-*144-4mve-en

! 4mpast44 *14ev-a t-t4mhed-or-a-e+eoh-twl-w444-4a4n-p4 *e o-4 n--t 46-pud

i TS.5.6-2 which u!11 have the capabl44ty te cheerb-energy ef- impaet-due te e ensk-drepr- Thie-*l-11 reeal t -- 1r. . m c4-reeter+1-de age tak4*g-plose-se 14ie-pect %!ch uculd recah-4n--el al41ecat t  ! c chage-f+em-th e-pwd,-

Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel assemblies.

C. Fuel Handling The fuel handling system provides the means of transporting and handling fuel from the time it reaches the plant in an uairtadiated condition until it leaves after post-irradiation cooling. The system consists of the refueling cavity, the fuel transfer system, the spent fuel storage pit, and the spent fuel cask transfer system.

Major components of the fuel handling system are the manipulation l crane, the spent fuci pool bridge, the auxiliary building crane, the j fuel transfer system, the spent fuel storage racks, the spent fuel l cask, and the rod cluster control changing fixture. The reactor vessel l stud-tensioner, the reactor vessel head lifting device, and the reactor l I internals lifting device are used far preparing the tsactor for refueling and for assembling the reactor after refueling.

Upon arrival in the storage pit, spent fuel will be reuoved from the transfer system and placed, one assembly at a time, in storage rccks using a long-handled manual tool suspended from the spent fuel pit -;

bridge crane. After sufficient decay, the fuel will be loaded into l

. a toragsTeaWQ foW s t6rigellin stiielIndependenRSpentf Fuel (S torage. ]

Just411atinnj rM ut6 shipping casks for removal from the site, The casks 1 will be handled by the auxiliary building crane.

T4:e Icad drap -ce nacquetre c c f-.-*--spent fuel cash fer Pr*4-r4+--4*knd hav. bec evclucted. It i c mat-pene t hl e , due-tc phye4 ec1 ccan,tw 4*4-a.,.

fer each te F^ dropped inte the 1 e r-ge.-f c e l (peci : 9  % Iced pat.h h m been definal-wh4eh-prev 4dec for cafe vecent e f.-tehc c c -%

  • revel 4*t+r-kekc-cnd : cchanical etcpc prevert cach avement rateide of th4e path. The-on1y aafc:> relatcd cquipment-that-cen-be-4mpsesed--dtree4y l during c ecck drep -cicag thic path-!^ ~he fac1 etcred-4n-the c=cil pae4-

! 9 001 a. 1). T'r concequ, c e c af-+hle-drop h~ r e b e c a e va l ua t e d a nd 4'e m! te acct th: NRC ctcff--er4-t e ric cant #hed-4n-42UREr 0512 i f as lecct 5 0 daya-4 Eve -ele p c ^ > c inee-r+aet+r-ekusdowef+r-f4*ef+n-g*o relence ccac4 der +shne-end-t4:e pcc1 unte r centche-et-lenet lo00 ppe boven-fe r c r i t ! c c l i ty r e n c 14e**t4c r - Uh44+44-days una determ hed adap*c tc , - -!*4au.- de ccy per4c d c f 5 ye e r 'cc heen-4m+per " '-" ^

L (4*ese- -teehnical - spee-141ca t ic _ t c provide-eddt-s.icrcl = rg4*-4e e t t i ng i 14ie-er i t e r ! - cpecif4ed i: """ EG-0613-fer-f4*s4en-gas-re4 e a c a c , uh44e -

net rectricting the pl:,t'- operatican! fla4bil i t y . ^

cask-4mpact l i-i t ^ r ^ r c rc ch--pad-prevent-e-*4 t ukf4eant-struetural--damage te the-pel 44eer,-

Spsnf fdelfdaskslui1E be; handled 1byTa? singleJ failure proof. handling;syntem ,

meeting (th'eFrequiremsnesfofLSection.5.1;6 ofLNUREG 061212" Controll oflHeavy Loadg at.J uclear: Power l.Pli' ants % July l1980.J Thefauxiliary;bui_1dingicrane hysjiiepn"l upgraded) to" conf 6rni'sith the singisifailure proof &equirosuht{of Se'etiont5.1;6ToflNURgG-0.612. ,The ? auxiliary; building cranel:1.s(designed lto n6t1lal16sf a'lloadjdrop asJa1 result'ofiany'aingle? failure. J Thelimproved l reliability of theiau.xilisry~buildihg) f crane is achieved through1 increased l factors of safety;and[through redundancy lor..dualityJin certain active copponents; i

1 l

Wb r6-4 Th cpe:et-f+m4-easi-w144--be-4eur red C6 fcet-f+om--4he-eu*l44e+y4wk4 ding te-4,he--rel4 read-ear---f+e-of4*14e-4+renoper4*t4eur----Epee 441 eat 4en-4 rS-nl44 44*14-thle-4eadleg-oper+t4e:; c ' ha t-i-E-+be-e*ek-drep: 65--feet, there wl44-not4Wgut 44eant-- : e l e e e r ef-f4eelen-prwheet+-f-rem-thMue4- 4c the-easky D. Spent Fuel Storage Cariacity The spent fuel storage facility is a two compartment pool that, if completely filled with fuel stort.ge racks, provides up to 1582 storage locations. The southeast corner of the small pool (pool no. 1) also serves as the cask lay down area. During times when the cask is being used, four racks are removed from the small pool, With the four storage rackt in the southeast corner of pool 1 removed, a total of 1386 storage locations are provided. t To allow insertion of a ehi+ying speiit[fliel cask, total storage is limited to 1386 assemblics, not including those assemblies which can be returned to the reactor.

Efference

1. USAR.-Section 10.2 1

l l

B.3.8-1 3,8 REFUELING._AND GEL HANDLING Bases The equipnent and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in interlocks and safety features., provido assurance that no incident could occur during CORE ALTERATIONS that would result in a hazard to public health and safety (Reference 1).

Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumenta-tion. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The resicual heat removal pump is used to maintain a uniform boron concentration.

Under rodded and unrodded conditions, the K g of the reactor must be less than or equal to 0.95 and the boron concentration must be greater than or equal to 2000 ppm. Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained.

3.8.A.1.h allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

No movement of fuel in the reactor is permitted until the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to permit decay of the fission products in the fuel. The delay time is consistent with the fuel handling accident analysia (Reference 2).

The cpent. fuel acced44er ill tw-4caded ir+: t he-+pmaMue-bonek ahr--ouR4e4+nt dcccy cf- fi=12- pr+ heter-4;h44+ -Imme(4*g--eud withdr-swlag the cach int +-poc1 "- 1, the-each u!!! be-enspen<k*! chc=

da bet t e cf-t4:e pc21 ep te c wieum-of--43-4eetuho-conempmuwo of pctent4+1 1 cd drepe--hn c bec: c vahated-in--aeeer-deswk4,

"""EC 0612 (Reference Q 911ew4ng i c c d1+eumlen-of-t4*e-4+aale-4ee 'j the-14*144t ie r uhleh-resul t c d f re+-that- evakat4ent The-eask-wil 1 nc t be-4m>er-t+1-4nt-e-4he--pool-+mt44-okMwbetw+ed-4n 4+-peobhas-4+en-414ebwgal-4+em--t4useeeter--*-+ek ! m= c f-bye <wer Gupper tin; anclya ke-Ewifeat+4-that-fuebot+ red ! n th*-tw,*bf+r-e P'+r i e d cc che r r cc 50 dcy: usu14-atlew-euf44el+nt deccy of-4he ficcice predec te cuc' that thel: relecm u:mid-w oul t ' n-of4--+44e dcce: 1:_ cc thca-26 t a f- 14*e-44 C" W r t 100 guide 4kom-The fS year desay--pet 4cd cc ncleetcd 'r fc14+w1*g thc gcner+bpr4*e41 4+--that opent-khitF t5, 1engcct "cecy time-wou14-++outt 1: thc lenoo

ff ,it
decec i: thc evcrt ' f :m-+ee4+k ut r-wh14e-prov14kg-tho-pbut-ope +4t4enel r1: n ib i444-yv Fuel lwill notLbe? ins _ertod'intola The spent fuel eask uill a t 4+-4mx+t+1 :r ui thdrewn-45*m--the-t+e4 un1ess a minin.um boron concentration of 1800 ppm is present. The 1800 ppm will ensure that 44-feo4-4e-ernobwn-by-a-eask i

l

B.3.8 2 3.8 REFUELING AND FUEL HANDLING Bnes continued herr k,,, fKtho?spppfu4ciu:k, lheluding!scatistf 6al Lun~certaintiesii' 1

-will be less than or equal to 0.95 fo tal11 postulated arrangementafofJfuel N13 M I K 5 N b Th; ^ Ch 2!Il ' be I mr+mi-**w4thdree-f+,m-the-peo+

u:lece : :: k-4 paeb-14mit-or r--erweh-telr-etwomb4emt4+n-the r e f l a-4n ple:: el44v-4.be-+opah444tt-4+-chser4+-eneref-+-eeek-dr+p-euch-44 *b-ee e4pM4eant-emeent f un:e r 1eekage-seeu14+--f4+m-peel-st,rcewrel--damager-

@dr i n t e ans*w: that s t- 2  != w441-w t e: Icvel drep 4*elew-4he-t.op-ef (4*e ;ent fuel :cred-4n-44a+-pool, In-4ece 4ng-44*e-ea ak--intmwr-elory 444eec--!: p e t e n t4el-dr+p-ef-45 fee t . Tk  ::5 u!4 4 ne h k wied-ent+

the-ear-r4*,r--for-eh4 ment-prinr t - Oc : 3 me nt-h-st-er+ge-pe r l e d . ^t--t-hle-44 mer the-+ed4eaet4+&+y-4 c de::y e d-co-that-e-relesee-ef-f4eekm-predm+t e-f res  ;

a44-fuel 23 :=h44+e-4c the ca" reuld-resu14,-4+-4M-sit + -deses-ken-14wn i 40-cm-peet-MO.--4 t !  : curred , f+r-+h4 : dear snelyci t hat--14-aseemh14+a cupwre-eft +r-eterege-for--M--daye-r-Othe : accuaptlena--or+-the e  :

t,heee-used-in-t44e-4:ppad fue1 aasembl-y- :::i dent-4c-+he-GER r--Seet4en-44r -

The recultant--denes e t t5: SIT" PDUNDARV-ere % Smo-t ' he - thyrol4-ond-L Rem-whole-bodyr The number of recently discharged assemblies in Pool No. 1 has been limited to 45 to provide-assurance that in the event of loss of pool cooling capability, at least eight hours are available under worst case conditions to make repairs until the onset of boiling.

The Spent- Fuel Pool Special Ventilation System (Reference 3) is a safeguards system which maintains a negative pressure in the spent fuel enclosure upon detection of high area radiation. The Spent Fuel Pool Normal-Ventilation System is automatically isolated and exhaust air is drawn through filter modules containing a roughing filter, particulate filter, and a charcoal filter before discharge to the environment via one of the Shield Building exhaust stacks. Two completely redundant trains are provided. The exhaust fan.and filter of each train are she ed with the corresponding train of the Containment In-service Purge Systen. High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers in each SFPSVS filter train. The charcoal adsorbers are installed to reduce the potential release of radiciodine to the environment.

During movement of irradiated fuel assemblies or control rods, a water level of 23 feet is maintained to provide sufficient shielding.

i.

l l

l t

t.

Bra,4-4 4,3-MFWM P" AMD - WR14ANDMNG -

Litteri-cent 4ma+4 The water level may be lowered to the_ top of the RCCA di've shafts for s

, latching and unlacching. The water level may also be lowered below 20--

feet for upper internals removal / replacement. The basis for these allowance (s) are (1) the refueling cavity pool has sufficient level to allow time to initiate repaits or emergency procedures to cool the core, (2) during latching / unlatching and upper internals removal / replace-ment the level is closely mohitored because the activity uses this level as a reference point. (3) the time spent at this Icvel is minimal.

The requirements for the storage of low burnup fuel in the spent fuel pool ensure that the spent fuel pool will remain suberitical during fuel storage.

Fuel stored in the spent fuel pool will be limited to a maximum enrichment of 4.25 weight percent U 235. It has been shown by criticality analysis that the use of the three out of four storage configuration will assure that the Kg, will remain less than 0.95, including uncertainties, when fuel with a maximtun enrichment of 4.25 weight percent U 235 and average assembly burnup of less than 5,000 MVD/MTil is stored in the spent fuel pool.

The requirement for maintaining the spent fuel pool _ boron concentration greater than 500 ppm whenever fuel with average assembly burnup of less than 5,000 MWD /MTU is stored in the spent fuel pool ensures that Egg -for the spent fuel pool will rernain less than 0.95, including uncertainties, even if a fuel assembly is inadvertently inserted in the empty cell of the three out of four storage configuration.

I i

l l

Re fe renG.ti!

1. USAR, Section 10.2.1.2

,. 2. USAR, Section 14.5.1 l 3. USAR, Section 10.3.7 kv-Ewhn i t C, "GP-bhne-AmenhetueM-O*4+d-DembF--34 r-4#4r i

<,w e.-, , - - ~ -

B.4.19-1 4(19 3 xiliary Butid_ inn crano~Liftint' Devices 10.12.A

)

Th'efcasiliary: build _inf erane has; been'modifiedito conform with: the' single

~

failure proof Lt equireminta _of Section: 5,116_' of: NUREG4612, " Control; of Heavy; Loads,at.hucioar. Power Planta",7 J ulyc1980. The auxiliary building crane isidesigned to not' allow a: load dropfasia resule of any-_ single x

failure, i As' the slings 'and_ special'lif ting / devices are by their nature, an11ntegral-part of theload bearing path, thetrisurveillance is.:nocessary ~

to ensure 7against:a' load droy as a. result _of' deficient rigging. Any'lo'ad that; weighs. niore. than- the cambined weight; of La single fuel ansemblyLand itsi associated handling ; tool is considered a bcavyl load.

G

___.____._m.-__-.m..___ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _

Exhibit C

- Prairie' Island Nuclear-Generating Plant:-

License Amendment Request Dated October 4, 1991 Revised Technical Specification Pages ,

- Exhibit C,. consists of the revised pages for the Prairie Island Nuclear.

Generating-Plant Technical Specification with the proposed changes incorporated. The revised pages are listed below: -

TS iil TS-vi TS-vii-TS xi TS,3.8 3 TS,4.1 24 (page -1 of 2)

TS,4.1 2B_(page 2 of 2)-

TS.4.19-1 TS.S.6 1 TS.S.6 2:

- B.3.8 1

'_ - B.3.8 2 B.4.19-1 D

i l

i.

-- ~ < , - - ,-r -y_,g v----- +w+.- ,-,+ce --

e +y,-w +,i,y g- e, e vr r - v-,-wey - - , .., e m -n re -e

-- g ..

. f 's TS-iii t

, - -TABLE OF CONTENTS (Continuedi TS'SECTION' -IIILE 'PAGE 3.6 Containment' System TS.3.6-1 '

A. Containment ~ Integrity

.TS.3.6-1  ;

B. Vacuum Breaker System TS.3.6-l C, Containment Isolation Valves 'TS.3.6 1 i D, Containment _ Purge System _ _

TS.3,6 E, Auxiliary Building Special Ventilation Zone Integrity .

TS.3.6-2 F.' Auxiliary Building Special Ventilation System TS.3~.6-3 C, Shield Building Integrity TS.3-6-3 H. Shield Building Ventilation System - TS.3.6 3 ,

I. Containment Internal Pressure TS.3.6-3 J. ' Containment and Shield Building Air Temperature TS.3.6-4 K. Containment Shell Temperature TS.3.6-4 '

L, Electric. Hydrogen Recombiners TS.3.6-4

.M. Containment Air Locks - TS.3.6-4 3.7. Auxiliary Electrical System TS.3.7 .1. 8 - Refueling-and Fuel Handling. TS.3.8 A.- Core Alterations TS.3.8-1 B. Fuel llandling Operations _

TS.3.8-3 l C.-Small Spent Fuel Pool Restrictions -TS.3.8-4

_D.. Spent Fuel Pool Special. Ventilation System TS,3.8 l E. Storage of Low Burnup Fuel TS.3.8 4-L3.9 Radioactive Effluents TS.3,9-1

-A, Liquid Effluents TS.3.9 1

1. Concentration TS.3.9-1 -

2.-Dose _

TS.3.9-I

3. Liquid Radwaste System TS.3.9 2 4.' Liquid Storage Tanks TS.3.9 2 B.' Caseous Effluents TS.3,9-3

,- 1,-Dose Rate TS.3.9 -2. Dose from Noble Cases fTS.3.9-3

3. Dose from I-131, Tritium and Redioactive Particulate' TS.3.9 4. Gaseous Radwaste Treatment System and Ventilation Exhaast Treatment-Systems TS.3.9 4 ,
5. Containment Purging 'TS.3.9-5 C. Solid Radioactive Waste TS.3.9-6 D. Dose from All Uranium Fuel Cycle Sources - TS.3.9-6 E. Radioactive Liquid Effluent Monitoring
  • TS.3.9-7 Instrumentation

! F.-Radioactive Gaseous Effluent Monitoring Instrumentation TS.3.9 7 l

l l

l r

7m-- q -%J

TS vi IA]iLE. .Q}' CONTENTS (qpntiaved)

T_S SECTIOS IIILE _l1L.,

4.12 Steam Generator Tube Surveillance TS.4.12-1 A. Steam Generator Sample Selection and TS.4.12 1 Inspection B. Steam Generator Tube Sample Selection TS.4.12-1 and Inspection C. Inspection Frejuencies TS.4.12-3 D, Acceptance Criteria TS.4.12 4 E. Reports TS.4.12-5 4.13 Snubbers TS.4.13-1 4.14 control Room Air Treatment System Tests TS.4.14-1 4.15 Spent Fon1 Pool Special Ventilation System TS.4.15-1 4.16 Fire Detection and Protection Systems TS.4.16 1 A. Fire Detection Instrumentation TS,4,16 1 B. Fire Suppression Vater System TS.4.16-1 C. Spray and Sprinkler Systems TS.4.16 3 D. Carbon Dioxide System TS.4,16-3 E. Fire llose Stations TS.4.16-3 F. Fire liydrant ilose llouses TS.4.16 4 C Penetration Fire Barriers TS.4.16 4 4.17 Radioactive Effluents Surveillance TS.4.17-1 A. Liquid Effluents TS.4.17-1 B. Gaseous Effluents TS.4.17 2 C, Solid Radioactive Waste TS.4.17-4 D. Dose from All Uranium Fuel Cycle Sources TS.4.17-4 4.18 Reactor Coolant Vent System Paths TS.4.18 1 A. Vent Path Operability TS.4.18-1 B. System Flow Testing TS.4.18-1 4.19 Auxiliary Building Cruno Liftiny, Devices TS.4.19 1 l

TS v11 1ABLE OF CORIEllTS (CDDilhutiO.

TS SECTION 1,11LE _ 6GE_

5.0 DESIGN FEATURES TS.S.1 1 5.1 Site TS,5.1-1 5.2 A. Containment Structures TS.S.2 1

1. Containment Vessel TS.5.2-1
2. Shield Building TS.S.2 2
3. Auxiliary Building Specini Ventilation Zone B. Special Vent 11ation Systems TS,5.2-2 C. Containment System Functional Design TS.S.2 3 5.3 Reactor TS.S.3-1 A. Reactor Core TS.5,3-1 B. Reactor Coolant System TS.5.3 1 C. Protection Systems TS.5.3 1 5.4 Engineered Safety Features TS.5.4 1 5.5 Radioactivei Wante Systems TS.5.5 1 A. Accidental Reicases TS.$ 5-1 B. Routine Releases TS.$.5 1
1. 1.iquid Wastes TS.5.5 1
2. Caseous Wastes TS.5,5 2
3. Solid Wastes TS.S.5-3 C. Process and Effluent Radiological Monitoring TS.S.5-3 System 5.6 Fuel llandling TS.$.6 1 A. Criticality Consideration TS.5.6-1 B. Spent Fuel Storage Structure TS.S 6-1 C. Fuel llandling TS.S.6 2 D. Spent Fue) Storage Capacity TS.S.6 2 l

%y .

TS xi 2 TABlj! OF CONTENTS (continuedl

!TS BASES SECTIQHe

_ TITLE ,_l PACE 4,0 BASES FOR-0URVfI LANCE REQUIREMENTS 4.1- Operativaal Safety Review E.4.1 1 4.2 Inservice Inspection and Testing of Pumps 'B.4.2-1:

and Valves Requirements 4.3 -Primary Coolant System Pressure Isolation- B.4,3-1 Valves 4,4 Containment. System Tests B.4.4"1 4.5: --Engineered Safety Features B,4.5-1

4. 6 = -Periodic Testing,of Emergency Power Systems B,4,6-1:

4;7- Main'Stesm Isolation Valves B.4.7 1-

.4.8 Steam and Power Conversion Systems- B.4,8 1 4.9f Reactivity Anomalies B,4.9 1 4.10 ' Radiation Environmental Monitoring Program B.4.10 1 A. Sample Collection and Analysis- B.4.10 1 B. Land Use Census- B.4.10-1 C,;Interlaboratory Comparison Program B.4.10 1

.4.11, Radioactive Source Leakage Test B,4.11-1 4.12 Steam Generator Tube' Surveillance B.4,12 1

' 4 : '13 Onubbers

. B.4.13 1 4.14. Control Room Air. Treatment-System Tests B 4.14-1

~

4.15 Spent Fuel Pool Special Ventilation System B,4.15-1 4,16 Fire Detectf.on and Protection Systems B.4,16-1

-4.17 Radioactive Effluents Surveillance B.4.17-1

=4.18 Reactor Coolant Vent System Paths B.4.18 1 4.19 ' Auxiliary Building Crane Lifting Devices B.4.19-1 l

4 l

l -.

l f.-

.c '

O  ;

s  !

TS.3.8 3 l

l 3,8,B. Fuel finndline. Operations i

1. 1/uring fuel handling operations the following conditions shall he satisiled:
a. Radiation lovels in the spent fuel storage pool area shall bo ,

monitored continuously during fuel handling operations. '

b. Prior to fuel handling operations, fuel handling cranes shall be load tested for OPERAb!LITY of limit switches, interlocks .

I and alarms,

c. A minimum boron concentration of 1800 ppm shall be maintained in the spent fuel pool whenever a spent fuel cask containing fuel is located in the spent fuel pool.

2 If any of the conditions in 3.8.B.1, abavn, cannot be met, sunpend all fuel handling operations and initinto the actions necessary to re establish compliance with the requirements of 3.8.B.1. '

?

C y% gT"TMF"--=*VF".my*-Tv.W wF-v--rge-w -w a)+*sg g19 9 q '=hv. ST-WPT-490-*'ip-m'p-p--r9m v-irerve *etang w % mSmew e wfTwevyit.y phyw g w+sy*vy MS-'

tF Mo 9Adup stNw9fMetw'*9

Tabic TS.4.1 2B (Page 1 of 2) l TABLE TS.4.1 2B  ;

tiUllMUM TRLOl'ENCIES P0h SAMPLINC TESTS FSAR Section TEST FREQUENCY Reference l

1. RCS Cross $/ week Activity Determination i
2. RCS Isotopic Analysis for DOSE 1/14 days (when at power)

EQUIVALENT.I 131 Concentration

3. RCS Radiochemistry E determination In mont.hs(l) (when at power)
4. RCS Isotopic Analysis for lodine 'a) Once por 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever Including 1 131, 1 133,'and 1 115 the specific activity ex-  ;

cceds 1.0 uCi/ gram DOSE ,,,

EQUIVALENT 1 131 or 100/E uCi/ gram (c or above cold shutdown), and b) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following TilERMAL POWER change exceeding 15 percent of t.ho RATED TilERMAL POWER within a ono hour >

period { above hot shutdown)

5. RCS P.adiochemistry (2) Monthly
6. RCS Tritium Activity Weekly 7 RCS Chemistry (Cl* , F* , 02) 5/ Week
6. _ RCS Borrn Concentration *(3) 2/ Week (4) 9.2
9. RWST Boron Concentration Weekly
10. Boric. Acid Tanks Boron Concentration 2/ Week

-11. Caustic Standpipe !?D11 Cnneentration Monthly 6.4 12, Accumulator Boron Concentration Monthly 6

13. Spent Fuel Pit Boron Concentration Monthly (7) 9.5.5 l 1

- w - , , - _ _ _

._...,5._._.__.-.a...__ . . _ . . _ . _ . - - _ . - _ . _ __ _ . - _ _ __ -

. 2 Table TS.4.1 28 (Page 2 of 3)

TABLE TS.4.1 ?B MINIMUM TREQUENCIFS FOR SAMPLING TESTS j FSAR Section [

TEST _ FREOUENCY _RefereD L i

14. Secondary Coolant Cross Weekly. l Beta Camma activity
15. Secondary Coolant Isotopic 1/6 months (5) .

Analysis for DOSE EQU1 VALENT

!.131 cencontration j

16. . Secondary Coolant Chem stry pl{ 5/ week (6)  !

pil Control Additive 5/ week (6)

Sodium 5/ week (6)

Notes:

F

1. Sample to be taken af ter a n.inimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was 14.st suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
2. To determine actisity of corrosion products having a half life greater than 30 rninutes.
3. During REFUELING, the borca concentration shall be verified by chemical analysi- daily.
4. The maximurn interval between analyses shall not exceed 5 days.
5. If activity of the samples is greater than 10% of the limit in Specification 3.4.D. the frequency shall be once per month,
6. The maximum interval between analyses shall not exceed 3 days.
7. The minimum spent fuel pool boron concentration from Specification 3.8 B.1.b shall be verified by chemical analysis weekly while a spent fuel cask containing fuel is located in the spent fuel pool.
  • - See Specification 4.1.D l

l l

l

. - e d

TS.4.19 1 4 . "' Auxiliary Buildinc.Ctane Lifting _ Devices AgIllD1h111LY Applies to surveillance requirements for the auxiliary building crane special lifting devices and slings before handling heavy loads carried over safe shutdown equipment or spent fuel in the spent fuel pool.

Objective To verify that special lif ting devices and slings used in conjunction with the auxiliary building cranc are operable prior to their use in supporting heavy loads over safe shutdown equipment or spent fuel in the spent fuel pool.

Speci!icatign Slings and special lifting devices which will be used in supporting heavy loads from the auxiliary building crane shall be visually inspected and verified OPERABLE within 7 days prior to their use in handling heavy loads over sinfe shutdoun equipment or spent fuel in the spent fuel pool.

. , s s TS.$.6 1 5.6 FUEL ilANDLING A. Criticality Considera1h D The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class I (meismic) structures. The spent fuel pit has a stainless steel liner to ensure against loss of water (Reference 1).

The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations.

l The design of the new fuel storage pit'and racks (Reference 1) ensures a ,

of less than or equal to 0.95, including uncertainties, .l new even fuel if unborate pit K,,,d water were used to fill the pit. The new fuel rack .!

coniiguration also ensures K , less than or equal to 0.98, including  ;

uncertainties, even if the n,ew, fuei racks were accidentally filled with a  ;

low density moderator which resulted in optimum low density moderation conditions. Fuel stored in the new fue. storage racks will have a maximum enrichment of 4.25 weight percent U 235.  :

The-spent fuel storage rack design (Reference 1) and the limitations on  !

the storage of low burnup fuel contained in Technical Specification Section 3.8.E ensure a spent fuel pool K,,, of less than or equal to 0.95, including uncertainties. The maximum enrichment of fuel to be stored in '

the spent fuel pool will be 4.25 weight percent U+235.

Fuel will not be inserted into a spent fuel cask in the pool, unless a minimum boron concentration of 1800 ppm is present. The 1800 ppm will ensure that k,, for the spent fuel cask, including statistical i uncertainties,,will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask. The criticality analysis for the TN.40 spent fuel storage cark was based on fresh fuel enriched to 3.85 weight percent U+235. i B. Spent Fuel Storare Structure  !

L

l. ,

The spent fuel storage pool is enclosed with a reinforced concrete building having 12 to 18-inch thick walls and roof (Reference 1).

The pool and pool enclosure are Class I (weismic) structures that afford protection against loss of integrity from postulated tornado  ;

missiles. The storage compartments and the fuel transfer canal are connected by fuel transfer slots that can be closed off with .

l~

pneumatically sealed gates. The bottoms of the slots are above the

. tops of the active fuel in the fuel assemblies which will be stored vertically in specially constructed racks.

The spent fuel pool has a reinforced concrete bottom slab nearly 6 feet thick and has been designed to minimize loss of water due to a dropped cask accident. Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel l

assemblies, s l

l l

o l,

e - e TJ,5,6 2

! C, fml _llandlitig The fuel handling system provides the means of t/ansportinn and handling fuel from the time it reaches the plant in an unitradiated condition until it leaves after post irradiation cooling. The system consists of the refueling cavity, the fuel transfer system, t he spent fuel storage pit, and the spent fuel cask transfer system.

Major components of the fuel handling system are the manipulation crano, the spent fuct pool bridge, the auxiliary building crane, the fuel transfer nystem, the spent fuel storage racks, the spent- fuel cask, and the rod cluster control changing fixture. The reactor vessel stud tensioner, the reactor vessel head lif ting device, and the reactor internals lifting device are used for preparing the reactor for refueling and for assemblin6 the reactor after refueling.

Upon arrival in the storage pit, spent fuel will be removed from the transfer system and placed, one assembly at a time, in storage racks using a long handled manual tool suspended frorn the spent. fuel pit bridge crane. After sufficient decay, the fuel will he loaded into storage casks for storage in the Independent Spent Tucl Storago Installation or into shipping casks for removal f rom t he ult e. The casks will be handled by the auxiliary building crene.

Spent fuct casks will be handled by a single failure proof handling system meeting the requirements of Section 5.1.6 of liUREG-0612, " Control of lleavy Loads at 11ucicar iower plants", July 1980. The auxiliary building crane has been upgraded to conform with the single failure proof requirements of Section 5.1.6 of liUREG 0617. The auxiliary building crane is designed to not allow a load drop as a result of any single failure The improved reliability of the auxiliary building crane is achieved through increased factors of safety and through redundancy or duality in certain active cotiponentS.

D. Epent Fuel StorarL Qipn i.17 The spent fuel storage facility is a t.wo compartment pool that, if completely filled with fuel storage racks, provides up to 1582 storage locations. The southeast corner of the small pool (pool no. 1) also serves as the cask lay down area. During times when the cask is being used, four racks are removed from the small pool. With the four storage racks in the southeast corner of pool I removed, a total of 1386 storngo locations are provided, To allow insertion of a spent fuel cask, total storage is limited to 1386 assemblies, not lueluding those assemblics which can be returned to the reactor.

Referet1CE

1. USAR, Section 10.?

o a o B.3.8 1 3.8 EfrUEL1FG AND PUEL,,jiA,NDLING ERLCA 5

The equipment and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built in interlocks and safety features, provide assurance that no incident could occur during CORE ALTERATIONf, that would result in a hazard to public health and safety (Reference 1).

Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumenta-tion. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.

Under rodded and unrodded conditicns, the K,,, of the reactor must be less than or equal to 0.95 and the boton concentration must be greater than or equal to 2000 ppm. Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained.

3.8.A.1.h allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

No movement of fuel in the reactor is permitted until the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to permit decay of the fission products in the fuel. The delay time is consistent with the fuel handling accident analysis (Reference 2).

Fuel will not be inserted into a spent fuel cask unless a minimum boron concentration of 1800 ppm is present. The 1800 ppm will ensure that k for the spent fuelcash,includingstatisticaluncertainties,willbel,e,ss than or equal to 0.95 for all postulated arrangements of fuel within the cash.

The number of recently discharged assemblies in Pool No. I has been limited to 45 to provide assurance that in the event of loss of pool cooling capability, at least eight hours are available under worst case conditions to make repairs until the onset of boiling.

The Spent Fuel Pool Special Ventilation System (Reference 3) is a safeguards system which maintales a negative pressure in the spent fuel enclosure upon detection of high area radiation. The Spent Fuel Pool Normal Ventilation System is automatically isolated and exhaust air is drawn through filter modules containing a roughing filter, particulate filter, and a charcoal filter before discharge to the environment via one of the Shield Buildirg exhaust stacks. Two completely redundant trains are provided. The exhaust fan and filter of each train are shared with the corresponding train of the Containment In-service Purge System. High efficiency particulate absolute (llEPA) filters are installed before the charcoal adsorbers to prevent clogging of the lodine adsorbers in each SFPSVS filter train. The charcoal adsorbers are installed to reduce the potential release of radiciodine to the environn.ent .

o eo B.3.8 2 3.8 REPUELING AND FUEL lbNDLING ]

Bases continued

-During movement of irradiated fuel assemblies or control rods, a water  ;

level of 23 feet is maintained to provide sufficient shielding.

The water level may be lowered to the top of the RCCA dri /o shaf ts for latching and unlatching, The water level may also be lowered below 20 feet for upper internals removal / replacement. The basis for these allowance (s) are (1) the refueling cavity pool has sufficient level to -

allow' time to initiate repairs or emergency procedures to cool the i core, (2) during latching / unlatching and upper internals removal / replace-ment the level is closely monitored because the activity uses this level as a reference point, (3) the time spent at this level is minimal. ,

The requirements for the storage of low burnup fuel in the spent fuel pool I ensure that the spent fuel-pool will remain suberitical during fuel storage. -t Fuel'atored in the spent fuel pool will be limited to a maximum enrichment of 4.25 weight percent U 235, it has been shown by criticality analysis that the use of the three out of four storage configuration will assure that the K will remain less than 0.95, including uncertainties, when fuel with a maxi,m,um enrichment of 4.25 weight percent _ U 235 and average assembly burnup of less than 5,000 MWD /MTU is stored in the spent fuel pool. ,

The requirement for maintaining the spent fuel pool boron concentration greater than 500 ppm whenever fuel with averago assembly burnup' of less than 5,000 MWD /MTU.is stored in the spent fuel pool ensures that K for the spentfuelpoolvillremainlessthan0.95,includinguncertaftties,evenif

  • a fuel assembly is inadvertently inserted in the empty cell of the three out ,

of four: storage-configuration.

r i

L F

? i k

References t

L 1. USAR, Section 10.2.1.2

2. USAR, Section 14._5.1 3, USAR, Section 10.3.7

,,__.1 _-. .--.m..-, U. . , , . , _ . . , . , , ., . u- ,....,....4._,..,,, ..y_,m.,r.. .m e ,.,_,.,.,..,_%...,_..., ~ . . , , - .,.m. ,....h_

c'e s*

11.4.19 1 4,19 Auxiliary Buildinn cranc Liftinn Dtvices Itases The auxiliary building crane has been modified to conform with the single '

failure proof requirements of Section 5.1.6 of NUREG 0612, " control of i lloavy Loads at Nuclear . Power Plants", July 1980. The auxiliary building i crano is designed to not allow a load drop as a result of any singic failure. As the slings and special lifting devices are, by their nature, an integral part of the load bearing path, their surveillance'is necessary  ;

to ensure against a load drop as a result of deficient rigging. Any load that weighs more than the combined weight of a singic fuel assembly and its  :

associated handling tool is considered a heavy lond, I t

+

I i

l b

P t'

s F

e t

4 I

9 i.-

4

,,,-..,d.m, . . , . - . . . . _ _ - - . . , _ . _ - - . . . , . . . . , . _ . - . . - . , - , . . . - . . . . . . . . _ _ . . . . - _ .._... -, _ _ .-- - - - .... -....~.,.._-.-.. ~ .4_

Retrieved from "https://kanterella.com/w/index.php?title=ML20085C999&oldid=2740320"