ML20045C171
| ML20045C171 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/11/1993 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20045C138 | List: |
| References | |
| NUDOCS 9306220153 | |
| Download: ML20045C171 (39) | |
Text
. . . _- = . . - ..
Exhibit B Prairie Island Nuclear Generating Plant License Amendment Request Dated June 11, 1993 Proposed Changes Marked Up On Existing Technical Specification Pages t
Exhibit B consists of existing and new Technical Specification pages with the proposed changes highlighted on those pages. The existing and new pages- i affected by this License Amendment Request are listed below:
TS-iii
-TS-vii TS-xiii TS.3.3-1 TS.3.8-4 TS.3.8-5 Figure TS.3.8-1 Table TS.4.1-2B (Page 1 of 2)
Table TS.4.'l-2B (Page 2'of 2).
TS.S.3-1 TS.S.6-1 ,1 TS.5.6-2 TS.5.6-3 Figure TS.S.6-1 Figure TS.S.6-2 B.3.8-2 ;
B.3.8-3 B.3.8-4 f
9306220153 DR 930611 p ADOCK 05000282 rN PDR y
TS-iii RE4-ee 7/9fe2 TABLE OF CONTENTS (Continued)
TS SECTION TITLE PAGE 3.6 Containment System TS.3.6-1 A. Containment Integrity TS.3.6-1 B. Vacuum Breaker System TS.3.6-1 C. Containment Isolation Valves TS.3.6 1 D. Containment 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 G. 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 3.8 Refueling and Fuel Handling TS,3.8-1 A. Core Alterations TS.3.8-1 B. Fuel Handling 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. Sterege of Lc'. Eurnup $~iig e Fuel fj g $jyfly 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. Gaseous Effluents TS.3.9-3
- 1. Dose Rate TS.3.9-3
- 2. Dose from. Noble Gases TS.3.9-3
- 4. Gaseous 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. Dose 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
TS-vii REV 99 7/9/92 ,
TABLE OF CONTENTS (Continued) ,
TS SECTION TITLE PAGE 5.0 DESIGN FEATURES TS.S.1-1 5.1 Site TS.S.1-1 5.2 A. Containment Structures TS.S.2-1
- 1. Containment Vessel TS.S.2-1
- 2. Shield Building TS.S.2-2 !
- 3. Auxiliary Building Special Ventilation Zone B. Special Ventilation Systems TS.S.2-2 C. Containment System Functional Design TS.S.2-3 5.3 Reactor TS.S.3-1 A. Reactor Core TS.S.3-1 B.-Reactor Coolant System TS.S.3-1 C. Protection Systems TS.S.3-1 5.4 Engineered Safety Features TS.S.4-1 5.5 Radioactive Waste Systems TS.S.5-1 A. Accidental Releases TS.S.5-1 B. Routine Releases TS.5.5-1
- 1. Liquid Wastes TS.S.5-1
- 2. Gaseous Wastes TS.S.5-2
- 3. Solid Wastes TS.S.5-3 C. Process and Effluent Radiological Monitoring TS.S.5-3 System 5.6 Fuel Handling TS.5.6-1 A. Criticality-Consideration TS.S.6-1 B. Spent Fuel Storage Structure TS.S.6-1 C. Fuel Handling TS.S.6-2 D. Spent Fuel Storage Capacity TS.S.6-3}.
TS-xiii
=RE" 92 3/12/90 APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop.
Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 K83 ppypyyps1[Pp61[Urjyyppip#pgEyfipg]MyipMj $MpMpIRiggijjisEj)]
3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In-Leakage Rate Sifil Sp3hE!?6iUPE61?BifHi6d/FfaiMCliFEEsrE6sfaTC6117ts pBE E@2 ppsh M $ @l[ $ f{[C$ kset(yh @ effp@ M K( $ $g yhs d { igifyppp y 6.1-1 NSF Corporate Organizational Relationship to On-Site Op+ rating Organizations 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-Site Operating Group
TS.3.3-1 RD1 91 10/27/99 3.3 ENGINEERED SAFETY FEATURES Applicability Applies to the operating status of the engineered safety features.
Obiective To define those limiting conditions that are necessary for operation of engineered safety features: (1) to remove decay heat from the core in an emergency or normal shutdown situations, and (2) to remove heat from containment in normal operating and emergency situations. ,
'1 Specifications A. Safety Iniection and Residual Heat Removal Systems
- 1. A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200*F unless the following conditions are satisfied (except as specified in 3.3.A.2 below):
- a. The refueling water tank contains not less than 200,000 i gallons of water with a boron concentration of at least 4950 2500 ppm.
- b. Each reactor coolant system accumulator shall be OPERABLE l when reactor coolant system pressure is greater than 1000 psig.
l OPERABILITY requirea.:
(1) The isolation valve is open '
(2) Volume is 1270 120 cubic feet of borated water (3) A minimum boron concentration of 1900 ppm (4) A nitrogen cover pressure of 740 i 30 psig
- c. Two safety injection pumps are OPERABLE except that pump control switches in the control room shall meet the require-ments of Section 3.3.A.3, 3.3.A.4 and 3.1.A.1.d.(2) whenever the reactor coolant system temperature is less than 310*F*.
- d. Two residual heat removal pumps are OPERABLE.
- e. Two residual heat exchangers are OPERAELE, l
- Vaird until 20 F.P2Y
_- .- . - - . - .. - -- - = . .
TS.3.8-4 01
.D -PU. 1 A,/S 7. ,/0 0 3.8.C. Small Spent Fuel Pool Restrictions No more than 45 recently discharged assemblies shall be located in the small pool (pool No. 1).
D. Spent Fuel Pool Special Ventilation System
- 1. Both trains of the Spent Fuel Pool Special Ventilation System shall be OPERABLE at all times (except as specified in 3.8.D.2 and 3.8.D.3 below).
- 2. With one train of the Spent Fuel Pool Special Ventilation System inoperable, fuel handling operations and crane operations with loads over spent fuel (inside the spent fuel pool enclosure) are permissible during the following 7 days, provided the redundant train is demonstrated OPERABLE prior to proceeding with those operations.
- 3. With both trains of the Spent Fuel Fool Special Ventilation System inoperable, suspend all fuel handling operations and crane operations with loads over spent fuel (inside the spent fuel pool ,
enclosure).
- 4. The provisions of specification 3.0.C are not applicable.
E. Ster ~~ ^f Lc" Burnun ruel
- 1. The felleuing rectrictienc ch:11 apply uhenever fuel zith'en 2"cr ge accc=bly burnup lecc then 5,000 m.9/"TJ ic ctered ir the cpent fuel peel (cr. cept ac cpecif4cd ir 2.8.E.2 and 3.8.E.2 beleu):
a,-The berer cencentretier ir the cpent fuel peel -h:11 bc int Lned-greater ther er equal te 500 pp=, and
- b. Fuci .:ith en average ecce=bly burrup 1ccc ther 5,000 un97373 ch:11 net be ctered i: =cre than three eterage 1ccatione-+4
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f* tLe cendbDienc b! ?. .b.1.2 abeve ere neD =ct, veIify tb I tbc spans-Twi4 peel cterage configur tier =cetc the requirc= ente of cpecificatie: 3. 8. E.1. b ca i-sucpend all re ttenc involving the
=everent cf fuel in-the cpent fuel peci until the beren cencentretien ic inerenced te 500 pp= er greater ,
- 3. If the condi tiene ir 3. 8.b.1.b abere are net =ct, cucpend all cet!cnc invelving =cvc=cnt cf fuel in the cpent fuct peel, verif.y the cpent fuel peel berer cencentretier te be greater than er equal te 500 pp= and iritiate corrective cctienc. Mic-pecitiened fuel ecce=blice ch:11 be --ved te acceptabic Iccati-onc prier te the recumptier cf ether fuel =cre ent ir the cpent fuel pect.
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- b. Fuel' assemblies Oith 'acombindt16n ~6f:b'ursupaiid ^ initial enrichment in the restricted range,of Figure TS.3.8 I shall'be stored,in,accordance with Specification 1 6.A.1.d.,
c f If' the' req 0i' emehts r 'of'^3~8'.E!1"Ji~^dnd 3;8 E!1;b '&r~e^ n'ot" ser';
immediately initiate action to move _any_ noncomplying fuel assembly,to,an acceptable location.;
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M Tha~ spent' ftiel'" pool' bbrdn do'ncinttation 'shall' bd W1,800 ~pps when fuel assemblies with a combination of burnup and initial enrichment in the restricted range, of Figure TS.3.B;1,are stored in the spent fuel pool and a spent fuel pool verification has not been performed'since the last mbyeiient 6f
~
' " ' ' ' ^ ' ^ ' ' ' '
any fuel, assembly in the spent fuel,,pooll" b'; if tho' requirements lof *specifidation 3,8,E.T.a "are~applicabid and the spent fuel pool boron concentration is n,ot within ,its limit, then imme_diately:,
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- 2) Either" initiate action' t6 're^stor'e spint fail p4o1 b6 rod concentration to within its limit,or perform a, spent, fuel pool verification.;
cl IThsipioVisissf6lfgpsdiffEstippy3:0!Claij[iiOE" appl 16Able!
FIGURE TS.3.8-1 12000 11000 I
/
/
10000 -
/
/
/
UNRESTRICTED
/
/
/.e 8000 '
/
h v y l
m 7000
/
/
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m e w 6000 '
o /
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s' M 5000 '
A
/
/
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2 4000 /
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g f RESTRICTED
/
/
/
2000 1000 S.5 4.0 4.5 5.0 INITIAL NOMINAL U-235 ENRICHMENT (w/o)
FIGURE TS.3.8-1 Spent Fuel Fool Unrestricted Region Minimum Burnup Requirements
Table TS.4.1-2B'(Page 1 of 2) '
RE'? 99 7/9/92 TABLE TS.4.1-2B liTNIMUM FREOUENCIES FOR SAMPLING TESTS FSAR Section TEST FREQUENCY Reference
- 1. RCS Gross 5/ week Activity Determination ,
- 2. RCS Isotopic Analysis for DOSE 1/14 days (when at power)
EQUIVALENT I-131 Concentration
- 3. RCS Radiochemistry 5 determination 1/6 months (1) (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 I-131, I-133 and I-135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE _
EQUIVALENT I-131 or 100/E uCi/ 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 /> following thermal POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown)
- 5. RCS Radiochemistry (2) Monthly
- 7. RCS Chemistry (Cl*,F*, 02) 5/ Week
- 10. Boric Acid Tanks Boron Concentration 2/ Week
- 11. Caustic Standpipe NaOH Concentration Monthly 6.4
- 12. Accumulator Boron concentration Monthly 6
- 13. Spent Fuel Pit Boron Concentration Monthly /yyhhijGMU 9.5.5-
_s
Table TS.4.1-2B (Pega 2 of 2)
REV 99 7/9/92 TABLE TS.4.1-2B MINIMUM FREOUENCIES FOR SAMPLING TESTS FSAR Section
- TEST FREQUENCY Reference
- 14. Secondary Coolant Cross Weekly Beta-Gam a activity -!
- 15. Secondary Coolant Isotopic 1/6 months (5)
Analysis for DOSE EQU1 VALENT I-131 concentration
- 16. Secondary Coolant Chemistry ,
pH 5/ week (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 determine activity of corrosion products having a half-life greater than 30 minutes.
- 3. During REFUELING, the boron concentration shall be verified by chemical analysis 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 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.
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TS.S.3-1 RE? 90 8/28/89 5.3 REACTOR A. Reactor Core
- 1. The reactor core contains uranium in the form of Kat@ggydslightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 pg(2pRp0jtubing to form fuel rods. The reactor core is made up of 121 fuel assemblies. Each fuel assembly contains 179 fuel' rods (Reference 1).
- 2. The maximum enrichment will be-445 gDjhreight percent U-235.
- 3. In the reactor core, there are 29 full-length RCC assemblies that contain a 142-inch length of silver-indium-cadmium alloy clad with stainless steel (Reference 2).
- 1. The design of the reactor coolant system complies with all appl-cable code requirements (Reference 3).
- 2. All high pressure piping, components of the reactor coolant system and their supporting structures are designed to Class I requirements, and have been designed to withstand:
- a. The design seismic ground acceleration, 0.06g acting in the horizontal and 0.04g acting in the vertical planes simultane-ously, with stresses maintained within code allowable working stresses.
- b. The maximum potential seismic ground acceleration, 0.12g, acting in the horizontal and 0.08g acting in the vertical planes simultaneously with no loss of function.
- 3. The nominal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet.
C. Protection Systems The protection systems for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1968. The design includes a reactor trip for a high negative rate of change of neutron flux as measured by the excore nuclear instruments (Reference 4). The system is intended to trip the reactor upon the abnormal dropping of more than one control rod (Reference 4). If only one control rod is dropped, the core can be operated at full power for a short time, as permitted by Specification 3.10.
References
TS.S.6-1 RE". o. o. '. ,' o. ,' o '
5.6 FUEL HANDLING A. Criticality Consideration E c ner and cpent fuel pit ctructurec crc derigned te uith: tend the enticipated certhquake Iceding: 2: C12cc 1 (ccic=ic) ctructurer. h c cpent fuel pit hoc : ctninlecc ctec1 liner te encure againct lecc cf unter (Reference 1)
.The neu and Ope"t fuel cterage r ch: cre decigned cc that it ic impeccibic te incert ess ablice ir ether than the preceribed lecatienc.
E c decign cf the neu fuel cterage pit cnd r chc (Reference 1) encur.es-a nw-f.uc1 pit Y,1, ef Iccc then er e ;ucl te 0.95, including uncertainticc, ever if unberated unter ucre ured te fill the pit. hc neu fuci rack
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37 Fuel will not be inserted into a spent fuel cask in the pool, unless i a minimum boron concentration of 1800 ppm is present. The.1800 ppm will ensure that k.rt for the spent fuel cask, including statistical 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 cask was based on fresh fuel enriched to 3.85 weight percent U-235. ,
B. Spent Fuel Storare 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 pool and pool enclosure are Class I (seismic) 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 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 assemblies.
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C. Fuel Handlina The fuel handling system provides the means of transporting and handling fuel from the time it reaches the plant in an unirradiated 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 crane, the spent fuel pool bridge, the auxiliary building crane, the fuel transfer system, 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 lifting device, and the reactor internals lifting device are used for preparing the reactor for refueling and for assembling 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 from the spent fuel pit bridge crane. After sufficient decay, the fuel will be loaded into storage casks for storage in the Independent Spent Fuel Storage ,
Installation or into shipping casks for removal from the site. The casks will be handled by the auxiliary building crane.
Spent fuel casks will be handled by a single failure proof handling system meeting the requirements of Section 5.1.6 of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", July 1980. Theauxiliarybuildingcyarne has been upgraded to conform with the single failure proof requirements of Section 5.1.6 of NUREG-0612. The auxiliary building crane is designed to not allow a load drop as a result of any single failure. The improved i reliability of the auxiliary building crane is achieved.through increased factors of safety and through redundancy or duality in certain active components.
i D. Spent Fuel Storane Capacity The spent fuel storage facility is a two-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 1 removed, a total of 1386 storage locations are provided. To allow insertion of a spent fuel cask, total storage is limited to 1386 assemblies, not including those assemblies which can be returned to the reactor.
Reference
- 1. USAR, Section 10.2 2'L 1"CritilEaliff7AhslyiE6EtlisEPfhlfibT1515hdilI61ts!1W2?Ffis1FahdESpsint@"Fds1
. ~
' RAcksQWejtlyhfusysymWNM1Mu$ lear 1Fd@DiyiAiknyFebrds$f39phg
FIGURE TS.S.6-1
". !!dni!!!! lH I
$N!!SO f!!j!!n!!!E!!!!!! i!!N.l$j miinna naiiniti !!!!!! tin if di!!!!!!i! !!!!!!i"!l !!!!!!!!!!!
H .dilii!!!! iii!!!!!!!: 1:...ili!!: i ki!!NN .!!i!!!N!N !!!!N!bb h hinHy lHMO R!!!!!!!!! !!!!!gitii
... ?!!!!!nii in!!!!!ii! !!nin!ni 131 iiiiiiiiii ii!!!!!!!H iiiiiiiiHi
'!I li!ENN!! !!!!!!N!!! !!!N!!!Ni
'!!!!!!!!! !!!N!!!!!! !!I!bb!N !!!!!!!!!!
)!!iiiiiiij iii!!!!!il! !!!:!!!!!ii iiiiiUili[
Siiiiiii; iiiiiiiiii! :i!!iiiiiii !!!!!iiiii:
.!! 9i!!!!!!!! !!!!Hi!!!! !!Hi!!!?!!
lj!jg jjlljjjiii !iiiiiHii! !!!!!iW!!
F;ijdS. iisiE!!N !!!!!E!!!! !Ei!!!E!!
' Eihemi- iiiniiiii! !!i!!!!!!!! ji!!!!!!!!!
!!!!!!!!!E ii!!!!!!!E ii!N!!Ui!
ill.ii,!!!!:
)!fh
, allii. iiii!iiiiii illililiiii iiiW" 8 '
=nmn; E
- m n iglitrm
- !!!!!!!;:i!
!!!!ii4"El :lii!!!!!j ,
h 8
!! kNi PATTERN FOR CHECKERBOARD REGION
- MMMMMMMM :
M W9XMMMMM '
"" f VA M M 11 MMM :::::ni:: .
!!! .:m
- iiki!!!i liiiiiiin iiniiiWi ,
N,k*,:iI i!O!!!!!! '
~
59l#h is!H!!il ..
/ -
wm- : :m: k .
' /'//f' i iiiisi!
kill!!ik[ "~'
Siii m;iiiii .
!!!!!!!@ p/I I
I BOUNDARY BETWEEN CHECKERBOARD AND UNRESTRICTED REGIONS l l
! l Fresh Fuel: Enrichments up to 5.0 w/o U-235, no restrictions on tureop l l l
jjjij=j Checkerboard Region 1
- i i Burned Fuel: Most settsfy minimun burnup requirements of Figure TS.5.6-2. )
I i
y Unrestrcited Region j
- Burned Fuel: Must satisfy minimum burnup requirements of rigure is.3.8-1. l 1
l Note: The Checkerboard and unrestricted regions can alternatively be separated by a single row of vacant cells on cach adjacent face.
FIGURE TS.5.6-1 Spent Fuel Fool Burned / Fresh Checkerboard Cell Layout I
I
FIGURE TS.S.6-2 30000
/
/
/
l /
l /
ACCEPTABLE /
/
n /
k20000 /
R /
- /,/
h so /
/
g 15000
)
/
f u
y / NOT ACCEPTABLE
/
I /
3 10000 7
)
/
=
/
/
\/
3.5 3.0 3.5 4.0 4.5 5.0 INITIAL NOMINAL U-235 ENRICHMENT (w/o)
FIGURE TS.S.6-2 Spent Fuel Fool Checkerboard Region Minimum Burnup Requirements
B.3.8-2 REU 99 7/9/93 3.8 REFUELING AND FUEL HANDLING 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 drive shafts for l 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 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 requircrcnte fer the cterage of leu burnup fuel ir the cpent fuel-pec1 ensure that the cpert fuel peci vill rc= ir cuberitical during fuel cterage, Fuel ctered ir the cpent fuel peel vill be limited te raxi=ur enrichment of ^ . 2 5 .:cight pe rcent U - 2 3 5. It hac been cheur by criticality en:1ycic that the uce of the three cut cf fcur cterage cenfiguratier vill accure that the M,t ui!! rem ir 1ccc than 0.95, including uncert Inticc, " hen fuel zi4h a-aa"*- ^nrich= cat cf ^ .25 ucight percent U-235 nd everage ecccably busnup cf 1ccc than 1,000 m3/MTJ ic ctered ir the cpent fuel peel.
The requirement fer reint:ining the cpent fuel peel berer cencentretien greater then 500 pp: "here zer fuel uith everage accc=bly burnup cf Iccc then 5,000 m3/MTJ ic ctcad- ir the cpent fuci peel encure: that Y,1, fer the
- pent fuel peel 2111 re m ir 1ccc then 0.95, including uncertainticc, even if a fuc1 ecce=bly ic inadvertently incerted ir the cepty ecil cf the three cut cf feu: cterage configuratier The Prairie' Island spent fuel' storage rects have~ been' analyzed"(P.eferencelh) t6 allow for the storage of fuel assemblies,with enrichments up to 5.0 veight percent U-235 while maintaining V 3 n s 0.95 including uncertainties. This criticality analysis utilized the following storage configurations or region's to ensure that the spent fuel pool will remain subcritical during the storage of fuel assemblies with a11]possible _ combinations of ,burnup and , initial enrichment:
li The first region utilizes m'checkerb'oard' loading' pattern'to' accommodate new or low burnup fuel with a maximum enrichment of 5,.0 wtt U 235. This configuration stores " burned" and " fresh" fuel assemblies in a'2x2 checkerboard pattern. Fuel assemblies stored in," burned" cell locations must have an initial enrichment less than 2.5 wet U-235 (nominal) or satisfy a minimum bdiaup requirement. The use of empty cells is als'o "a'n acceptable option for the " burned" cell locations. Fuel assemblies stored in the " fresh" cell locations can have enrichments up to 5.0, wt( U ,23,5 vith
,no requirements for, burnup or_ burnable absorbers.
2J The "second~ regioxi'does not' utilize"any, speci'al' loading' pattern ~. '" Fue,1 assemblies with burnup and initial enrichments which fall into the
' nrestricted range of Figure TS.3.8t1 can be stored anyWhere in ths~' region u
with no special placement restrictions.' Fuel assemblies which fall'into the restricted range of Figure,TS.3.8-1 must be stored in,t,he checkerboard region in accordance with Specification 5.6.A.1.d.
ES..89 3.8 REFUELING AND FUEL HANDLING Bases continued ThdKharWEd/frsshi f6elWh'sEkhrE6sidfrshiO6?Esh"belpusitf3Wsd MhywhsrWVithih?ths
- s pe,n.ttfs61T,fAc1&,sN,bsd.x.~nA%6nda'rpit/A.tssenienUhh.s.ekelb6AfdN,e.gion.
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u.. nr.e,. s t r i c.t- e- d i.re gi on i m.-.u.s t. .e be~{ _ . xeither:i 1K.aspsiata.dl?bys.N.aushti.W._ou. '6.fTs.
. - - -- s11@n.6s thorsN E 22n cth,ss,'in,tsi.:.faEFTmUnt,.Ebs?E6nfigu.
ss w
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- =3
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~ --
FiguisETS73!83T@h1Ehispsi;ilfishTthsi'siihiiiGiiibuYhupTis 61fssisduiffiti hnresthictidNtofag~ef,_&235
~ inffthnsp'snQd
- n. ~. n .+ a cw e nEnrich,il? pod 1RI.Al6A ~ ~ ~.a . n ~~
to i 5.0~we ight 7perce.~nt.- . mentsM10wer ,btha ti
~.
conservatively lb.oundedsbysthe0 minim.;.w umib,u_rnu~p=i .
- .e.
re qu i r_n;3i,~8~71we,i ementif6r$3;87 n -
te v.,ight ;
~
ght percenttU;235,fwhich,fisl20,00,f NVD/MTUh
. - . , .w. IThe.refoi,~sW,.Figske1TSf3.a!,8si3.fasXbei.dT_drasa
~- - ~
to requireithstJfue1%~thfantinitia10enribhm~ent?ofilessithanb318RGeight i
ilA r. c. ent_%2351hA.ve520. 00!M. O. D/M.. TU$buf.
.- -h.briMMs.reAidribA_fb..rA?shrs_s_tY1_E_ES_.C.
-- ;. A ._s isthe! spentsfielipoo_U . ~ . - ~ ~ .
s_ill? be ; a, l.lowe'di.
The water' in' the~ spent fder pool 'normally "contains soluble' boron. ~which~ results
~
in large suberiticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the' 11miting k.tr' of 0.95'be evaluated in the absence of soluble boren,"Hence, the design of both regions is based on the use of unborated water, which ensures' that each region is ~ '
maintained in a suberitical condition during normal, operation with,the regi6ns ful. lLsl.o. adede - -
r ifiEsht*?ihEfasistin3ha Most?ssbidsutN6ndi~tioniTd6'~hste ac t iviteo f feiEhe r % fs thh tso @egibus@Ma$@le sidit91hMsf@MfithhAeiddihditcond arei thhtlasslof?c6oiingdshifdfdpping$fla?fuh19nsssnblyZdnthbditshi6ffthe ~ ~ ~
haEk kackh@Asilhsfdr8pfingsaf?
desig
+. .
preclude sn ithis c c6ndi dien%d a'?Niif.~ass cw HoNeveh esis.%sussEf9AeM nKacciMn@NaNM,pddt,0,I$tsd$,t cm v +
m$ddle sd co n,, ,s.w'r l-t..h...e.. .t d..
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- :- - /%
t..s.eo i...v. ity:.1 u-ef appliedh ?Fo, i r.(Thiif,stsatesithaffens t t.he. .:x <lacc,s,d, ;-
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'is<notqrequireld'toJassumeitwa?uniik$lyhindipsndenti
- e. nsur. e .p'.r'o.S- te.d.. tis. hT. A Iins. tM. ~esit._iiss.li. ^ ~~
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Theid6ubleRonti6gs6dfspiih61pTsTs116ssNrsdi M fdds F10bTs~b6f6K"undsFabh6fiiid1 orf ace'idsntF c6nditisns MAinc6fon1MaNingiefsOfdsti neid%6 fednsiidsrsd? Atsone ~ ~^
tide.7toEekampledtneimB$d misihAd,ingi,,ofM.fdeEnkied1%yisversnesidsndscenasiMidEtNNaas
- - . ~ an s.- a nd6N x w[N.u.N c wu~ n,~$atihn %or N. n - - ..- ~
ori lo. .. cation.
criticality
,, . .~ a,n.iastiMt.isidv 2related?AccidentsM spent.n. .
xw ifi~ cation ~23 f B iE s2 Spec c c. ;
~b at!eliN6ssij_dt!61,stE_d"
.a:a Fensures t th=eispentsfusi ,ilYraEksi ST. . -
poolicontAitis EAde q6stid$lis s sIYdd SofonianfEideffds fNAAAmbI1$ANiihNi ~~~~
cnbisun16nifMurnuplandiihifisi T.S2328MneYSi,.bFidyid.
- b. . ,, a .een performed'isi;: ce_,
n . < -tthR1,a. s
^
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~. t? move m entTo' fi$.Yp.p4nes tfde,is,ypW61Mi.
n; . , . , .,/ < w : - s f.th. sN, de1#.;;p;fensiEn y... .,: . nL~. '-
sapent2 fuel ' tfica
. .n m poot 2 Theinksativ$; reac. tlYityteffscE6f? he,isolubles<bbeohSb61dicompehsdreff6E t
~ + c ,. . - - -- -
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i i
l w
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l 4
I 3.8 REFUELING AND FUEL HANDLING l r
Bases continued TbsTU6rbh7h'oii6shtirktf6h7EsijdirsmssEsEofiSps'difiEsti'6h?3!8!E72Istsss5"[16 hist isposes%s'n7rMfdelied0ese6EsMEsfoddushihgG5d Mn hy{ bye $;cdmM1hh%bjdanslef fdyejifINyjop(3]BjE7 and S16MY11dfarejthenladequateft,olpreventferftipalityg t@(thrj$yrsMifmdn Specification '3.8.E.2,a is 'not imposed Nhen~ only' fu~el" assemblies" with 'a combination of burnup and initial enrichment in the untestricted range of Figure TS,3.8-1 are stored in the spent fuel pool. The requirements of Specification 3.6.E.2.a are not required in that case because with only' fuel assemblies that have burnup and initial enrichment in the unrestricted rangefof Figuro TS.3.8 1 it is not possible to cause an inadvertent critical,ity,by mispositioning,a fuel, assembly,in the spent fue,1, pool g Whss*thE?iequifesshes"6f]Sp6Eifial:f6ni3!&2Eq273?ifE!ipp11&b1Myshdiths caconc ensration tion:lmds fo {fjt61 ti bel taken bofos[ ins th'el(s p re cluden ppt~ence thef occurs @fue1 { po61$i i o CanYa silis sj thsihl req 61@its c c identfot thefconse'quences[of{sn:(jecidenEin3065resshyfhll s jIlsido;s$fflyi4$hly$ehisisd by:-; imme dila e elyys usp ending[theJnovlese nte fffuelj[as shnb11ss j .jThs sio negtra;t 16n of boroniis) restored?simultaneouslyWithisu' spending;movementioftfdel asseablies) :(;inj{ acceptable;%1tbrnativeitsitofb6mplete3afspents:fus1?pa61 verificatiad :(However#psieriltidiieiiusin5fsovssistjofffael?hsshmbli.es@ia co.ncentrariostofiborb.iisust%
+.:u > .
elf;ressor.s.at iThisTd.6esinotip~ruc.indsilmoveinentI6fFi
- - - - - - e -+ - - - -
fddl l a s s emblylto;;;gs a fe~posl ition; A"spehtiffseUp65FysrlfEati;o;hfis:11sidir'sWTo115pingths!Tssinii6ysnish@fishf fuel + assembly?infthef:spentifuel?po61$1fifdel?nsssmblissNith7A2combinstias?6f bdrnup(andliriitial]estichsssty inYthsifs:ssfidtedjfsussi:L6fMidilrs!TSf 318 Elf srs ~
s tored ithithel spe nt? fus1Tpoo1MThis svb rifinasi6n ikillic defisa fthat?snylf6el ~
hssnablies!sithfaysombinstionGfibninsp?and?iditisite6flihshusfiFthf ~
ke s tEistei range} $ f ;;Fighfe :LTS ;;3 i8$1;/;;arcksEojedlinidec6 fdsne d%ftMthd requireme'nts ibff Spydif.ihation[5s6 Al;.1.dj l
l References
- 1. USAR, Section 10.2.1.2 !
- 2. USAR, Section 14.5.1
- 3. USAR, Section 10.3.7 4T. '
- Crit icali tyl Analfili!b fl the;i Praliis? Island [ Units [1TT Ff e shTand ] SpEntg:Fud17 ^ _
- Rasks * ,3es tinghosseJCommercial; N&cle ari Fdelj Diji nion @ February [19932 ;
i i
1
t i
Exhibit C +
i Prairie Island Nuclear Generating Plant. !
License Amendment Request Dated June 11, 1993 -
Revised Technical Specification Pages-i Exhibit C consists of revised and new pages for the Prairie Island Nuclear ;
Generating Plant Technical Specifications with the proposed changes :
incorporated. The revised and new pages are listed below: '
i TS-lii TS-vii ,
TS-xiii TS.3.3-1 !
TS.3.8-4 TS.3.8-5 .
Figure TS.3.8-1 Table TS.4.1-2B (Page 1 of 2)
Table TS.4.1-2B (Page 2.of.2) '
TS.S.3-1 TS.S.6-1 :
TS.5.6-2 .!
TS.S.6-3 Figure TS.S.6-1 Figure TS.S.6-2 B.3.8-2 B.3.8-3 B.3.8-4 I
'l i
TS-iii 1 l
t TABLE OF CONTENTS (Continued)
TS SECTION TITLE PACE !
3.6 Containment System TS.3.6-1 _
A. Containment Integrity TS.3.6-1 -!
B. Vacuum Breaker System TS.3.6-1 C. Containment Isolation Valves TS.3.6-1 D. Containment 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 ;
TS.3.6-3 G. Shield Building Integrity 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 3.8 Refueling and Fuel Handling TS.3.8-1 .'
A. Core Alterations TS.3.8-1 ,
B. Fuel Handling 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. Spent Fuel Pool Storage TS.3.8-4 l .
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 r
- 3. Liquid Radwaste System TS.3.9-2 4.-Liquid Storage Tanks TS.3.9-2 B. Gaseous Effluents TS.3.9-3
- 1. Dose Rate TS.3.9 ~;
- 2. Dose from Noble Cases TS.3.9-3
- 4. Gaseous 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. Dose 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 f
b
TS-vii TABTS OF CONTENTS (Continued) ,
I
.TS SECTION TITLE PAGE 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 s
- 2. Shield Building TS.5.2-2
- 3. Auxiliary Building Special Ventilation Zone-B. Special Ventilation Systems TS.S.2-2 C. Containment System Functional Design TS.5.2-3 5.3 Reactor TS.S.3-1 A. Reactor Core TS.S.3 '
B. Reactor Coolant System TS.S.3-1 C. Protection Systems TS.S.3-1 5.4 Engineered Safety Features TS.5.4-1 '
5.5 Radioactive Waste Systems TS.S.5-1 A. Accidental Releases TS.S.5-1 B. Routine Releases TS.S.5-1 *
- 1. Liquid Wastes TS.S.5-1 .
- 2. Gaseous Wastes TS.5.5-2
- 3. Solid Wastes TS.5.5-3 C. Process and Effluent Radiological Monitoring TS.5.5-3 System 5.6 Fuel Handling TS.S.6-1 A. Criticality Consideration TS.S.6-1 B. Spent Fuel Storage Structure TS.S.6-1 '
C. Fuel Handling TS.S.6-2 D. Spent Fuel Storage Capacity TS.S.6-3' l b
s i
I
TS-xiii APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FTGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and liydraulic Two Loop.
Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-7 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant-Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 3.8 1 Spent Fuel Fool Unrestricted Region Minimum Burnup Requirements 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid ,
Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In-Leakage Rate 5.6 1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6-2 Spent Fuel Fool Checkerboard Region Minimum Burnup Requirements 6.1-1 NSP Corporate Organizational Relationship to On-Site Operating '
Organizations 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-Site Operating Group
'i l
l i
l I
. - ~ . - .~ . ~. .- .- .- .
TS.3.3-1 3.3 ENGINEERED SAFETY FEATURES Applicability Applies to the operating status of the engineered safety features. ,
Obiective i F
To define those limiting conditions that are necessary for operation of ',
engineered safety features: (1) to remove decay heat from.the core in an emergency or normal shutdown situations, and (2) to remove heat from ;
containment in normal operating and emergency situations.
Specifications A. Sa fe ty Iniection and Residual Heat Removal Systems 8
- 1. A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200*F unless the following conditions are satisfied (except as specified in 3.3.A.2 below):
- a. The refueling water tank contains not less than 200,000 L gallons cf water with a boron concentration of at least i 2500 ppm.
b.' Each reactor coolant system accumulator shall be OPERABLE when reactor coolant system pressure is greater than 1000 psig.
OPERABILITY requires: ,
(1) The isolation valve is open (2) Volume is 1270 1 20 cubic feet of borated water .
(3) A minimum boron concentration of 1900 ppm ;
(4) A nitrogen cover pressure of 740 30 psig
- c. Two safety injection pumps are OPERABLE except that pump control switches in the control room shall meet the require-ments of Section 3.3.A 3, 3.3.A.4 and 3.1.A.1.d.(2) whenever the [
reactor coolant system temperature is less than 310*F*.
c
- d. Two residual heat removal pumps are OPERABLE. i
- e. Two residual heat exchangers are OPERABLE.
i e
i
- Valid until 20 EFPY i
TS.3.8-4 ,
1 3.8.C. Small Spent Fuel Pool Restrictions No more than 45 recently discharged assemblies shall be located in the small pool (pool No. 1).
D. Spent Fuel Pool Special Ventilation System 1.
Both trains of the Spent Fuel Pool Special Ventilation System shall be OPERABL2 at all times (except as specified in 3.8.D.2 and 3.8.D.3 below).
- 2. With one train of the Spent Fuel Fool Special Ventilation System inoperable, fuel handling operations and crane operations with ,
loads over spent fuel (inside the spent fuel pool enclosure) are permissible during the following 7 days, provided the redundant train is demonstrated OPERABLE prior to proceeding with those operations.
- 3. With both trains of the Spent Fuel Pool Special Ventilation System inoperable, suspend all fuel handling operations and crane operations with loads over spent fuel (inside the spent fuel pool enclosure). .
4 The provisions of specification 3.0.C are not applicable.
E. Spent Fuel Pool Storare
- 1. Fuel Assembly Storage a, To be stored without restriction in the spent fuel pool, the burnup and initial enrichment of a fuel assembly shall be within the unrestricted range of Figure TS.3.8-1.
- b. Fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure _TS.3.8-1 shall be i stored in accordance with Specification 5.6.A.l.d.
- c. If the requirements of 3.8.E.1.a and 3.8.E.1.b are not met, immediately initiate action to move any noncomplying fuel t assembly to an acceptable location.
- d. The provisions of Specification 3.0.C are not applicable. }
1
- , - - . . m-- - - ..,.s ,,.,._.___m - .
TS.3.8-5 !
3.8.E.2. Spent Puel Pool Boron Concentration
- a. The spent fuel pool boron concentration shall be a 1,800 ppm when fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stocod in the spent fuel pool and a spent fuel pool fication has not been performed since the last movement of -
.uel assembly in the spent fuel pool,
- u. If the requirements of specification 3 8.E.2.a are applicable l and the spent fuel pool boron concentration is not within its limit, then immediately:
- 1. Suspend movement of fuel assemblies in the spent fuel pool, and
- 2. Either initiate action to restore spent fuel pool boron concentration to within its limit or perform a spent fuel pool verification.
- c. The provisions of Specification 3.0.C are not applicable. ;
i L
v i
4 t
{
I
FIGURE TS.3.8-1 12000 11000 i l
/
10000 .f
/
/
/
UNRESTRICTED /
g
/
/
8000 l p If g2
/
/
m 7000 /
O /
b
/
/
ro 6000 o f k
D j' 5 5000
/
/
A F l' b
- c 4000 f M
/
/
/
3000
, RESTRICTED
/
/
f
/
2000 1000 S.5 4.0 4.5 5.0 INITIAL NOMINAL U-235 ENRICHMENT (w/o)
I I
FIGURE TS.3.8-1 Spent Fuel Fool Unrestricted Region i Minimum Burnup Requirements !
l 1
Table TS.4.1-2B (Page 1 of 2)
TABLE TS.4.1-2B MINIMUM FREOUENCIES FOR SAMPLING TESTS FSAR Section~
TEST FREOUENCY Reference
- 1. RCS Gross 5/ week Activity ~ Determination
- 2. RCS Isotopic Analysis for DOSE 1/14 days (when at power)
EQUIVALENT I-131 Concentration
- 3. RCS Radiochemistry E determination 1/6 months (1) (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 I-131, 1-133, and I-135 the specific activity ex-ceeds 1.0 uCi/ gram-DOSE _
EQUIVALENT I-131 or 100/E uCi/ 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 /> following thermal POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) ,
- 5. RCS Radiochemistry (2) Monthly
- 7. RCS Chemistry (Cl*,F* , 02) 5/ Week
- 9. RUST Boron Concentration Weekly
- 10. Boric Acid Tanks Boron Concentration 2/ Week
- 11. Caustic Standpipe NaOH Concentration Monthly 6.4 I
- 12. Accumulator Boron Concentration Monthly 6 l
~
- 13. Spent Fuel Pit Boron Concentration Monthly / Weekly (7H8) 9.5.5 I
i 1
. . = . _ _ . . .- -
Table TS.4.1-2B (Page'2'of 2).
TABLE TS.4.1-2B i
MINIMUM FREOUENCIES FOR SAMPLING TESTS FSAR Section TEST FREOUENCY Reference
- 14. Secondary coolant Gross Weekly Beta-Camma activity
- 15. Secondary Coolant Isotopic 1/6 months (5)
Analysis for DOSE EQUIVALENT .
I-131 concentration
- 16. Secondary Coolant Chemistry pH 5/ week (6).
pH Control Additive 5/ week (6) '
Sodium 5/ week (6)
Notes:
- 1. Sample to be taken after a minimum of 2 EFFD 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 determine activity of corrosion products having a half-life greater than 30 minutes. .
- 3. During REFUELING, the boron concentration shall be verified by chemical analysis 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 maximum interval between analyses shall not exceed 3 days.
- 7. The minimum spent fuel pool boron concentration from Specification. l 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.
- 8. The spent fuel pool boron concentration shall be verified weekly, by chemical '
analysis, to be within the limits of Specification 3.8.E.2.a when fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in the spent fuel pool and a _ :
spent fuel pool verification has not been performed since the last movement of any fuel assembly in the spent fuel pool, t
- See Specification 4.1.D
TS.S.3-1 5.3 REACTOR A. Reactor Core
- 1. .The reactor core contains uranium in the form of natural or slightly l enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 or ZIRLO tubing to form fuel rods. Thereactorcoreismadel up of 121 fuel assemblies. Each fuel assembly contains 179 fuel rods (Reference 1).
- 2. The maximum enrichment will be 5.0 weight percent U-235.
- 3. In the reactor core, there are 29 full-length RCC assemblies that contain a 142-inch length.of silver-indium-cadmium alloy clad with stainless steel (Reference 2).
- 1. The design of the reactor coolant system complies with all appl-cable code requirements (Reference 3).
- 2. All high pressure piping, components of the reactor coolant.
system and their supporting structures are designed to Class I requirements, and have been designed to withstand:
- a. The design seismic ground acceleration, 0.06g acting in the horizontal and 0.04g acting in the vertical planes simultane-ously, with stresses maintained within code allowable working stresses,
- b. The maximum potential seismic ground acceleration, 0.12g, acting in the horizontal and 0.08g acting in the vertical planes simultaneously with no loss of function.
- 3. The nominal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet. l C. Protection Systems The protection systems for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1968. The design includes a reactor trip for a high negative rate of change of neutron flux as measured by the excore nuclear instruments (Reference 4). The system is intended to trip the reactor upon the abnormal dropping of more than or.e control rod (Reference 4). If only one control rod is dropped, the core can be operated at full power for-a short time, as permitted by Specification 3.10.
References
TS.S.6-1 5.6 FUEL HANDLING A. Criticality Consideration
- 1. The spent fuel storage racks are designed (Reference 1) and shall be maintained with:
- a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;
- b. K.tr s 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Reference 2; i
- c. New or spent fuel assemblies with a combination.of burnup and .;
initial enrichment in the unrestricted range of Figuro.TS.3.8-1 allowed unrestricted storage in the spent fuel racks; and
- d. New or spent fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 stored in compliance with Figures TS.S.6-1 and TS.S.6-2.
- 2. The new fuel storage racks are designed (Reference 1) and shall be maintained with:
- a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; l
- b. K.tr 5 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Reference 2; and
- c. K,gg 5 0.98 if accidentally filled with a low density moderator which resulted in optimum low density moderation conditions.
1
- 3. 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.tr for the spent fuel cask, including statistical 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 cask was based on fresh fuel enriched to 3.85 weight percent U-235.
B. Spent Fuel Storare 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 pool and pool enclosure are Class 1 (seismic) 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' 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.
TS.S.6-2 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 assemblies. i The new and spent fuel pit structures are designed to withstand the '
anticipated earthquake loadings as Class I (seismic) 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.
C. Fuel Handlinz The fuel handling system provides the means of transporting and i
handling fuel from the time it reaches the plant in an unirradiated 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 crane, the spent fuel pool bridge, the auxiliary building crane, the fuel transfer system, 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 lifting device, and the reactor internals lifting device are used for preparing the reactor for refueling-and for assembling 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 from the spent fuel pit bridge crane. After sufficient decay, the fuel will be loaded into storage casks for storage in the Independent Spent Fuel Storage Installation or into shipping casks for removal from the site. The casks will be handled by the auxiliary building crane.
Spent fuel casks will be handled by a single failure proof handling system meeting the requirements of Section 5.1.6 of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants", July 1980, The auxiliary building crane has been upgraded to conform with the single failure proof requirements of Section 5.1.6 of NUREG-0612. 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 components.
_ _ ,. == ;. - . _-
'TS.S.6-3 D. Spent Fuel Storare Capacity The spent fuel storage facility is-a two-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 1 removed, a total of 1386 storage locations are provided. To allow insertion of a spent fuel cask, total storage is limited to 1386 assemblies, not including those assemblies which can be returned to the reactor, f
I 9
'I 1
l l
Reference i
- 1. USAR, Section 10.2 I
- 2. " Criticality Analysis of the Prairie Island Units 1 & 2 Fresh and Spent Fuel l Racks", Westinghouse Commercial Nuclear Fuel Division, February 1993.
FIGURE TS.S.6-1 e n ilmam 10!!0!U taiina
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PATTERN FOR CHECKERBOARD REGION MMMBB VARM Interface MMM9V4%%M EBEER@ i %.2..W.
Wh Sih":ii-$lJi] un L gF4$$
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I BOUNDARY BETWEEN CHECKERBOARD AND UNRESTRICTED REGIONS l Fresh Fuel: Enrichments up to 5.0 w/o U-235, no restrictions on burnup l
in E Checkerboard Region d$"i'il Burned Fuel:
It Must satisfy mintran burnup requirements of Figure is.5.6-2.
' Unrestreited Region i 5 Burned Fuel: Must satisfy minimum burnup requirements of Figure Ts.3.8-1.
Note: The Checkerboard and unrestricted regions can alternatively be separated by a single row of vacant cells on each adjacent face.
FIGURE TS.5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout
FIGURE TS.S.6-2 l
30000
/
/
/
25000
/
/
ACCEPTABLE /
/
m c
/
20000
/
8
" /
/
P g
[
5
- f
/
g 15000
/
h u
/
NOT ACCEPTABLE
/
l10000
- c
/
M
)
/
5000
/
/
/
/
l h./
5 \ 3.0 3.5 4.0 4.5 5.0 INITIAL NOMINAL U-235 ENRICIDfENT (w/o) l l-l l
FIGDRE TS.5.6-2 Spent Fuel Pool Checkerboard Region Minimum Burnup Requirements
B.3.8-2 3.8 REFUELING AND FUEL HANDLING 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 drive shafts 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 Prairie Island spent fuel storage racks have been analyzed (Reference 4)-to allow for the storage of fuel assemblies with enrichments up to 5.0 weight percent U-235 while maintaining K.tr 5 0.95 including uncertainties. This criticality analysis utilized the following storage configurations or regions to ensure that the spent fuel pool will remain suberitical during the storage of fuel assemblies with all possible combinations of burnup and initial enrichment:
- 1. The first region utilizes a checkerboard loading pattern to accommodate new or low burnup fuel with a maximum enrichment of 5.0 wt% U-235. This configuration stores " burned" and " fresh" fuel assemblies in a 2x2 checkerboard pattern. Fuel assemblies stored in " burned" cell locations must have an initial enrichment less than 2.5 wt% U-235 (nominal) or satisfy a minimum burnup requirement. The use of empty cells is also an ,
acceptable option for the " burned" cell locations. Fuel assemblies stored in the " fresh" cell locations can have enrichments up to 5.0 wtt U-235 with no requirements for burnup or burnable absorbers.
- 2. The second region does not utilize any special loading pattern. Fuel assemblies with burnup and initial enrichments which fall into the unrestricted range of Figure TS.3.8-1 can be stored anywhere in the region with no special placement restrictions.. Fuel assemblies which fall into the restricted range of Figure TS.3.8-1 must be stored in the checkerboard region in accordance with Specification $.6.A.1.d.
The burned / fresh fuel checkerboard region can be positioned anywhere within the spent fuel racks, but the boundary between the checkerboard region and the unrestricted region must be either:
- 1. separated by a vacant row of cells, or
- 2. the interface must be configured such that there is one row carryover of the pattern of burned assemblies from the checkerboard region into the first row of the unrestricted region (Figure TS.S.6-1).
l l
l
. _ _ _ _ - _ _ - - _ _ _ _ _ _ - _ _ _l
_ . - . . . _. . _- - . ~ . _ . - - - .. . .
B.3.8-3 3.8 REFUELING AND FUEL HANDLING Bases continued Figure TS.3.8-1, which specifies the minimum burnup requirements for unrestricted storage in the spent fuel pool, is based on enrichments from 3.87 to 5.0 weight percent U-235. Enrichments lower than 3.87 weight percent are ,
conservatively bounded by the minimum burnup requirement for 3.87 weight !
percent U-235 which is 2000 MWD /MTU. 'Therefore, Figure TS.3.8-1 has been drawn to require that fuel with an initial enrichment of less than 3.87 weight percent U-235 have 2000 MWD /MTU burnup or greater before unrestricted storage in the spent fuel pool will be allowed, r The water in the spent fuel pool normally contains soluble boron, which results ;
in large suberiticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting k.rt of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both' regions is based on the use of unborated water, which ensures that each region is maintained in a suberitical condition during normal operation with the regions fully loaded.
Most accident conditions do not result in a significant increase in the activity of either of the two regions. Examples of these accident conditions are the loss of cooling, the dropping of a fuel assembly on the top of the rack, and the dropping of a fuel assembly between rack modules and wall (rack design precludes this condition). However, accidents can be postulated that could increase the reactivity. For these accident conditions, the double contingency principle of ANS1 N16.1-1975 can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.
The double contingency principle allows credit for soluble boron under abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is the accidental l misloading of a fuel assembly into a rack location for which the restrictions on location, enrichment or burnup are not satisfied. This could potentially increase the reactivity in spent fuel racks. To mitigate these postulated criticality related accidents, Specification 3.8.E.2 ensures the spent fuel pool contains adequate dissolved boron anytime fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in the fuel pool and a spent fuel pool verification has not been performed since the last movement of any fuel assembly in the spent fuel ;
pool. The negative reactivity effect of the soluble boron would compensate for !
the increased reactivity caused by a mispositioned fuel assembly.
The boron concentration requirements of Specification 3.8.E.2 are no longer ;
imposed when no fuel movements are occurring and a spent fuel pool verification has been completed, because the storage requirements of Specifications 3.8.E.1 and 5.6.A.1.d are then adequate to prevent criticality.
l Specification 3.8.E.2.a is not imposed when only fuel assemblies with a l combination of burnup and initial enrichment in the unrestricted range of l Figure TS.3.8-1 are stored in the spent fuel pool. The requirements of l Specification 3.8.E.2.a are not required in that case because with only fuel assemblies that have burnup and initial enrichment in the unrestricted range of Figure TS.3.8-1 it is not possible to cause an inadvertent criticality by mispositioning a fuel assembly in the spent fuel pool.
B.3.8-4 3.8 REFUELING AND FUEL HANDLING Bases continued When the requirements of Specification 3.8.E.2.a are applicable, and the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. An acceptable alternative is to complete a spent fuel pool ,
verification. However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.
A spent fuel pool verification is required following the last movement of fuel assemblies in the spent fuel pool, if fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in the spent fuel pool. This verification will confirm that any fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in accot/snce with the requirements of Specification 5.6.A l.d.
1
't s
1 I
References l i
- 1. USAR, Section 10.2.1.2
- 2. USAR, Section 14.5.1 j
- 3. USAR, Section 10.3.7
- 4. " Criticality Analysis of the Prairie Island Units 1 & 2 Fresh and Spent Fuel j 3 Racks", Westinghouse Commercial Nuclear Fuel Division, February 1993. l I
l
1 l
I
-l 2
I i
Exhibit D l Prairie Island Nuclear Generating Plant License Amendment Request Dated June 11, 1993 criticality Analysis of Prairie Island Units 1 and 2 -
Fresh and Spent Fuel Racks ,
Prepared By ,
Westinghouse Commercial Nuclear Fuel Division February 1993 t
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