ML20079L225

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs,Revising Updated Final Safety Analysis Rept for Extended Burnup Fuel
ML20079L225
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 10/30/1991
From:
HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20079L196 List:
References
NUDOCS 9111060174
Download: ML20079L225 (27)
v • d • e


Text

.

STPEGS UFSAR F ATTACHMENT 3 ST HL-/sE-39o&

PAGE I 0F M TABl.E 1,6 1.(Continued)

Fon-VESTINCHOUSE REPORTS INCORPORATED PY REFERENCE REPORT TITLE UFSAR REFERENCE NUREG/CR 1741 Models for Escination of Incapacication 2.2.3 Times Following Exposure to Toxic Gases or Vapors (December 1980)

TOP 2 Design for Pipe Break Effects, 3.6 Rev. 2 (May 1974) hibTOP-4 Subcompartment Pressure Analysis, 3.6 i Rev. 1 (October 1977_) __

-_p go gg.cx .Sooq A sresime# of Exdd L g d to L f Ll h V W k s h a ( & n uyIfflh II,l.h fl. 2.3 1 IF. A . 4

~

\

l 4

1.6-4 Revision 0 9111060174 91103o PDR ADOCK 05000498 PDR

ATTACHMENT 3 STPFCS UFSAR ST HL AE 390 6 PAGE_,LL__0F 3 7 3.11.5.2 . Radiation EnvironmeDI. Safety related systems and components are designed to perform their safety related functions after normal operation radiation exposure plus a DEA exposure. The normal operational exposure is based on the design basis source tarus presented in Sections 11.1,-11.2, 11.3, and 12.2.1 and'the equipment and shielding configurations given in Section 12.3.

Safety-related system and component radiation exposures are dependent on equipment location and the particular DBA involved. . In the Containment and "

control room area, equipment exposures are based on the DBA LOCA.- For in-Containment equipment, the DEA LOCA source term is based on a release of 100 percent of the core noble gases, 50 percent of the halogens and 1 percent of.

the solids. This is consistent'vith the guidance-given in RG 1.89. Control room exposures following a postulated IDCA( k nd : the-source-ee._ .cdf44d 4n- Sectice 15.5.5.0, are controlled to 5 rads or less consistent with the requirements-of GDC 19 of 10CFR50, Appendix A.

7ascar AL }

Radiation source terms for safety-related components which are exposed to post-accident recirculation fluid are consistent with the recommendations of RG 1.89 (i.e., 50 percent of the core halogen inventory and 1 percent of the remaining core solio fission product inventory are mixed in the recirculation water).

Normal.and accident radiation doses for the various plant areas are presented in Table 3.11 1. Safety related equipment design doses are the sum of normal plus accident exposures. The design radiation exposures delineated in Table 3.11-1 are based on gamma and bata radiation. Radiation source terms for

safety-related components outside Containment are based on gamma radiation.

, Organic materials in the Containaent are identified in Section 6.1.2. For the

organic coating materials-used inside Containment (see Section 6.1.2.1),

irradiation tasts performed by Oak Ridge National Laborarory have been performed for an integrated gamma dose of 1 x 108 rads (which exceeds the desigt, calculated value in Table 3.11-1). These doses conservatively account

-for the surf ace exposure due to beta radiation in the design basis LOCA environment.

3.11 5 Revision 0

ARACHMENT.3 ST HL AE 3*)o lo INSERT "Ala PAGE_ 1 _ 0F M _

The release pathways considered for the PBA lhCA are described in Section 15,6.5.3. .The source terins used correspond to a cycle length of 20,000 MWD /MTU,' a core average burnup of 40,000 MWD /MTU, and a discharge burnup of 60,000 MVD/MTU. These burnups are conservative relative to the planned cycle lengths described in Section 4.3. -,

-i l

i i

USQi91-287,001

TAELE 3.11-1 (Continued) .

DMIRONwE*TTAL CONDITIONS Relative Cumulative Radiation"8 Pressure Emidity Dosare --

Location Terroe ra ture Normal Range Accident Norinal Accident tS8 Radiation (Environmental Normal Range Abnormal' AccidentW (rads) TrDe Normal Ace 4 dant'38 (max / min. 11 ft) (rads)

Desirnator) (max / min. *F) (*F) 51 70/0 100 3.5x10'  ! . '. r 0 5 gamma Reactor"8 135/65 142/- 325 +0.3/psig

'l 6*X/u s e-beta Cavity max

-0.1 pr.ig 2.5x10** neutron (Ras 001, 002) min m_'

325 +0.3/psig 51 psig 70/0- 100 3.5x10 5 L ' -19 gamma Other Areas Inside('8 120/65 167/- #8K/C8 and max max Secondary Shield beta

-0.1 psig -3.1 psig (Below E1. 19 ft. min min Rs. 004) 325 +0.3 psig 51 psig 70/0 100 2x10' 1.2x10' gamma Other Areas Inside 120/65 165/- and max max Secondary Shield beta

-0.1 psig -3.1 psig (above El. 19 ft) min sin ,

Other Areas Outside 120/65 168/- 325 +0,3/psig 51 psig 70/0 100 3.5x10* W #

gamma max max /- 6ft O and Secondary Shield beta

-0.1 psig -3.1 psig sin min 325 +0.3/psig 51 psig 70/0 100 7x10 5 1.2x10' gamma RHR Pump and Heat 120/65 167/- and Exchanger Rooms; max max 5

-3.1 psis beta Valve Rooms (Ras. 104 -0.1 psig 109, 110, 301, 304, min sin 306, 105. 108, 111, 202, 209, 207) stu. 80/20 100 100 100 gamma Tendon Access Callery 95/50 120 BA sta.

(Pes. 011. 013)

M

\m n

O >

>s1 I m.m5

~d 3.11-7 oj u u

3 a

~ - - _ _ _ - _ _ _ _ _ _ _ _ _ _

mB12 3.11-1 (Continued) ..

DTVIROP* ENTAL CONDITIONS Relative Cumulative Radiation"3 Musidity Doser.e 1.ocation Terecerature Pressure Normal Range Accident Normal Accident m Raciation.

(Envirormental Normal Range Abnorm.a1 Accident8 (reds) Tme Normal Accident'88 (reax heln. 41 (41 (rads)

Designajor) (max / min. *F1 (*F) ata. 80/20 80 100 1.3x10' gamma HVAC Room 104/50 104/50 104 ata.

(Ras. 508, 509) 104 men, ats. 80/20 80 100 1.3x10' 6 *"*"

Radiation Monitor 104/50 104/50 (Rm. 510)

4. Elf) u ndline Buildint 120 slightly sligntly 80/62 100 10' l 2x10' garsse IWAC Supply 104/65 117/62 Subsystems) nigativa negative i

(Rm. 002) 120 slightly slightly 20/20 80 10 8 100 gamma

!GAC Room 120/65 120/62 negative negative (Rs. 003) 120 slightly slightly 80/20 60 10'  ? . h E" gansas ZCCS Cubielee (Ras. 004, 005, 006) 104/E,5 104/62 negativa negative F///O' 104/65 104/62 120 311ghtly slightly B0/20 20 10' W c gamme Recirculation valve negative M X/O Rocas (Ras. 007, negative 00s, 009) 120 clightly slightly 20/20 100 10' 2.1x10' gamma Spray Additive 104/65 120/62 Tank Rooms negative negative 4

(Rms. 007A, 008A, 009A) s116h tly 80/20 80 l ')' 7.L10* gammaa IWAC Room 104/65 107/62 120 slightly f.y aio, (Rm. 010) negativa negative 120 alightly slightly 80/20 100 10' IGO p HtAC Carbon 104/65 120/62 Filtsr Roce negative negative . p (Rm. ICI-)

,y 4p ISi>

I t W

!ME dw@

E* .;

oO y 3.11-13 , ." @

c( - ii 1 1 s_

( . .

I. .

g g g~ [ ~

SbHL AE390 sTTEcs Ursa PAGE_4.__.OF.22..-

u___ --

4.3.1.1 Fuel faImm.

f.1111 The fuel rod design basis la described in Section 4.2. The rr.telear dasign basis is *.o install sufficient reactivity in ths fuel to attalu a region average discharge turnup of approx 12aately M7900 MVD/MTU. The above, along with the design basis in Noction 4.3.1.3, control of Power Distribution, satisfies CX 10.

g Q 46 N Fuel burnup is a usasure of fuel depletf on which represents tha integrated energy output of the fuel (KVD/MTU) and is a convenient meant for quantifying fuel exposure criteria.

The core design lifetime or design discharge burnup is achisved by installing sufficient initial excess reactivity in sach fuel region and by following a fuel replacement program (such as .het descri'oed in Section 4.3.2) that rseste all safety related criterie in each cycle of operation.

Initial excess reactivity insts11ed in v.he fuel, a(though not a design basis, nust be sufficient to maintain core criticality at full power operating conditions throughout cycle life with equilibrium xenon, samarium, and other

-fission products present. The end of design cycle life is defined to occur when the chemical shin concentration is essentially taro with control rods present to the degree necessary for operational requirsments (e.g. , the cot. trolling bank at the " bite

  • position). In terms of chersical chim boron concentration this represents approxinetely 10 ppm with no control red insertion.

A limitation on initial installed excess reactivity is not required other than as is quantified in terns of other design bases tauch as core negative reactivity -feedback and shutdown margin discussed below.

4,3.1.2 Feettivr_.Raletivity rti& Ash (Reactivity Coeffisimul.

31111 The fuel teoperature coefficient will be negative and the moderator temperature coefficient of reactivity will be nonpositive for power operating conditions, thuahy providing negative reactivity feedback characteristics.

The design basis usets CDC 11.

D11IN.1Jil2D When compensatien for a rapid in:rease in reactivity is considared, there are tvo major effects. These are the resonance absorption affec,ta (Doppler) associated with changing fuel teeperature and the spectrum effect resulting frets changing mcderator density. These basic physics characteristics are often identified by renetivity coefficients. The use of slightly enriched uraniu:n ensures that the Doppler coefficient of reactivity te negativa. This coefficient provides the most rapid reactivity compensation. Tha core is aluo designed to have an overall negative moderator tercperature coefficient of renetivity so that average coolant temperature or void content provides another, slower cowpensatory effect. Nossinal power operation le permitted 1

4.' 2 Revision 0 l

e . .

ATTACHMENTI- '

ST HleAE-370 6

STFECS UrsAR PAGE '7 2 0F[7 -

oscillations can be excited by prohibitad setion of individual conersi rods.-

Such oscillationa are readily observable _ and alarmed, usin6 the excore long -

ion chambers. Indications are also centinuously cvailabla from incore thermoccuples and loop tawparature motsursments. Moveable incera desseters can be activated to provida arte datailed information. In all proposed cores these horizontal _ plans oscillations are self. damping by virtue of reactivity -

g feedback effects desigm d into the core.

(

However, axial menon rpatial power oscillations may occur during core life.

The control banks and excore detectors are providad for control and nonitoring of axial power distributions. A.ssurance that fusi design limits are not p exceeded-is provided by reactor overpower AT and overtemperatura AT trip g functions which use the measured asial power imbalanca as an input.

' 4.3.1.7 Anticinated Trgnaients Withggt Trie (ATVT). The effects of. -

anticipated transiants with failure to trip are not considered in the design bases ol' the pb.nc. Analysis has shown that the likelihood of such a -

hypothetical event is maligibly small (Ref. 4.31). Furthermore, analysis of

, the consequences of. a hypothetical failure to trip following anticipated.

'i transients has shovn that no significant core damage would result, system peak

! pressures vould be limited to acceptable valuas, and no failure of the Smactor

. Coolant System (KCS) would result, hse analyses were documented -(Ref.

U 4.3-2) in November 1974 in accordance with the Atomic Energy Commission'(AIC)

g. policy outlined in WASH 1270 ' Technical Report on Anticipatsd Transients j Vithout Scram for Vater-Cooled Power Reactors", september 1973.

i 4.3.2 Description

[ 4.3.2.1 Egglear Desien Descrie* ten. = h reactor core consists of a L specified number of fuel rods which are held in bundlas by spacer grida and L top - and betten fittings, he fuel rods ars constructed of Zircsloy L cylindrical tubes containing uranium dioxide fuel pellets.- The bundles. known .

as fuel assemblies, are arranged in a pattern which approximates a right, circular cylinder.

Each fuel assembly contains a 17 x 17 rod array composed of 264 fuel-rods, 24 l rod cluster control thimbles and an incora inarrumentation thimble. Figure L 4.2-1 shova a cross sectional view of a 17 x 17 fuel assembly and the related t rod cluster control locations.- Further details of the fuel assembly are given L in' Section 4.2.

L The fuel rods within a given assembly have the same uranium enrichment in both the radial and axial planes. Fuel assemblies of three different enrichments l' are used in the initial core loading to establish a favorable = radial power distribution. Firure 4.31 shows the fuel loading pattern to be used in the first core. Two regions consisting of the two lover enrichments are interspersed so as to form a checkerboard patterm in the: central. portion of the-core, h third region is arranged around'the periphery of the cora and contains tha highest onrichasnt, h enrichments for the first core are shown in Table 4'.3 1.

5 m reference reloading pattern i.s typically similar to Figure 4.3-1 with depleted fuci interspersed checkerboard style in the conter and new fuel mixed l~

vith depleted fuel on the periphery. The core will normally operate on a wo md l c t/d* **- Me: month cycle, accumu? eting approximately 10."lC HVD/MTU burnup. h

exact reloading pattern. initial and final positio of assemblies, number of 4.3 6 D# Revision 0 r-

,7

,y .i.

_m . ._. _ _

i gsWPtb- Q Y 4 , y N;Jl _.

y R,M &r*W ~

g ( _

_ q sTyrcs UrsAst '5hACHMENT 3 ST-HL-AE3M fresh ' assemblies and their placement kra wpp dependent 0F oLfhT.4MfF~

j' for th. next cycle and burnup and power histories of the travious cycles.8 OC#

The core average enrichment is detensined by the amount of fis.f onable l material required to provide the desired core lifetime and energy , /

requirements, namely-a region average siischerge burnup1f 3+r0901tVb/MTU. The physics of the burnup process is such that operation of the reactor depleten the amount of fuel available due to the absorption of neutrons by the uranium.

235 atoms and their subsequent fission. -The rate of uranium.235 depletion is directly proportional-to the power level at which the reactor is operated. In addition, the fission process results in the formation of fission products, some of which readily absorb neutrons. These effects, depletion and the <

U buildup of fission products, are partially offset by the buildup of plutonium shown on Figure 4.3 2 for the 17 x 17 fuel assembly, vbich occurs due to the t nonfission absorption of neutrons in uranium-238. Therefore, at the beginning '

of any cycle a reactivity reserve equal to the depletion of the fissionable fuel and the buildup of fission product poisons over the specified cycle life

, must be ' built' into the reactor. This excess reactivity is controlled by i~ removable neutron absorbing material in the form of boron dissolved in the

. primary coolant and in the essa of the first cycle, by burnable poison rods.

4 '

The concentration of boric acid in the primary coolant is varied to provide control and to compensate for long term reactivity requirements. The concentration of the soluble neutron absorber is varied to compensate for

reactivity-changes due to fuel burnup, fission product poisoning ' including

! xenon and samarium, burnable poison depletion, and the cold to operating moderator temperature change. Using its normal makeup path, the CYCS is j: capable of inserting negative reactiviry s.t a rate of approximately 60 pcm/ min l vhan the reai. tor coolant boron concentration is 1000 ppa and approximately 70

, pcm/ain when the reactor coolant boron concentration is 100 ppa. The peak .

4 burnout rate for menon is 25 pea / min (Section 9.3.4.3.1 discusses the j- capability of the CVCS to counteract xenon decay). Rapid transient rasetivity requirements and safety shutdown requirements are met with control rods.

,co Y As the boron concentration is increased, the moderator temp 9tature coefficient becomes less negative. The use of a soluble poison alone vedd result in a

positive moderator coefficient at beginning of life (BOL) for the -f+rst- cycle.

Therefore, burnable poison rods are used in the-44*ee-core to reduce the soluble boron concentration sufficiently to ensure that the moderator temperature coefficient la negative for power operating conditions. During

,p operation the poison content in th:se roda is depleted thus adding positive reactivity to offset sons of the negative reactivity from fuel depletion and i ' -

fission product buildup. The depletion rate -of the burnable poison rods is not critical since chemicalshin is always available -and flexible enough to ,

. I% cover any possible deviations in the expected burnable poison depletion rates

- Figure 4.3-3 is a graph of a typical core depletion with and without burnabic poison rods. Note -that even at end of life (EOL) conditions some residual poison remains in the burnable poison rods resulting in a net decrease in the  ;

--fitee-cycle lifetime. Upom-sepinie ef 2: firse-eyeksi'l-the burnable '

.- poison rods are swaselly ::r:4 hern the moderator tempa.,rature coefficient 1 rele d = u is sufficiently_ nag niv.a wl ca neerssm T. e t 4 2n addition to reactivity control the burnable poison rods ara strategie. ally located to provide a favorable radial power distribution. A'igure 4.3 4 shows the burnable poison distribution within a fuel asseobly for the several B

h 4.3 7 Revision 0 r  ;

3

ATTACHMENT 3 ST ML AE 390(o hg PAGE _1 0F ._d __

ds-f j' ,

STPICS UrsAR Pover distribution, rod ojection, and

  • rod misalignment atrlyses are based on j

~t he arrangement of ,the shutdown and control groups of the RCCAs abovn on

. Figure 4.3 36. All shutdown RCCAs are withdrawn befc.ro withdravel of the  !

control banks is initiated. In going frca zero to 100 percent power, control  !

banks A, 5, C, and D are withdrawn sequentially. De limits of rod p:,sitions  ;

and further discussion on the bacio for rod insertion limits are provided in the Technical Specifications. l 4.3.2.4.13 hann fedsnt.ItuttAhtta: Reactor coolant (or moderator) temperature control has added flexibility in reactivity control of the W s t inghous e WR, This featurs takes advantage of the negative moderator corporature coefficient inherent in s WR to:

1. Muisite return to power capabilities.
2. Provide 15 percent power loud regulation capabilitter without requiring control rod compensation.
3. Extend the time in cycle life to which daily load follow operations can be accomplished.

Eeactor coolene temperature control suppleaants the dilution capability of the plant by .tevering tha reacter coolant tenperature to supply positive

roactivity through the negative moderator tersperature coefficient of the i rea
tor. Af ter the transient la over, the system automatically recovers the reactot coolant tarperature to the programmed value.

Moderster temperature coattel of reactivity, lika soinbia boron control, has.

l' the advantage of not significe.ncly affecting the core power distribution.

However, un11ke boron cor.crol, tamperature control can be rapid encush to achieve reactor power t.bange races of 5 percent / minute.

4.3.2.4.14 Eurnable Peise Rods: The buxuble poison rods provide partial control ef the wxcoss reactivity available during the 4 hat fuel j; cycle. In doing so, these rods prevent the acderator temperature coefficient froo being positive at nonsal operating conditions. They perform thf.c function by reducing the requirenent for soluble poison in the moderator at the beginning of the first fuel cycle as described previously. For purposas of illustration a typical.burnabla poison rod pattarn in the core, togethag-with the innber of roda per caseably, is shown on Figura 4.3 5 while tha. Wdk arrangements within an aJaenhly are displayed au rigura 4.3 4.O'e'~rTect vorth of v.hese rode is shven in Tabla 4.31. ,is baron in the roda is deplet3d with buruup but at a sufficiently slow rate so that the resulting l

critical concentration of soluble boren 1o such that the mudarxtor temparature

( coefficient remains nogetive at all times for power operating conditions. ,

i 4.3.2.4.15 Zenk 7enon startum: . ccepensation for the peak xenon buildup L'

is a.:complished using the boron control systan. Startup ftote the peak xenon condition is accomplishR with a combination of rod mutian and boron dilution.

The boron dilution ny be made e.t any time, inc1witng during the shutdown period, provided 'ha shutduwn margin is nintainad.

l 4.3.2.4.16 Luihllev control..aniltn2a.'Janttd: During load follow maneuvera, p:ver changes are accomplished using control tod motten and l- dilution or boracion b;; the boren system as required. Control rod motion is limited by the contre.1 rod insertion limita in the Technical Specificar. ions 4,3-26 Revision 0 '

l l

' l AllACHMENT3 ST-HL-AE 390 4 STFICS WSAR PAGE n_0F EL_. _

11.0 P & l t 'IIX L VA$1 L n6E M L1rEl g 11.1 SOURCE Tr.pys Source tern:s and niodels used in the evaluation of radvaste treatment systems and effhent roles ses are based on operating plant data where ava11abit, (Ret. 11.1 1).

Two source terus are presented; desf gn basis and realistic. Design basis source terns for shielding design and component f ailures are based upon the sa.te censervative model for reactor coolant activity. Shielding source terns are discussed in core detail in Section 12.2. The source terum for the affluent release analysis are based on the realistic model for reactor coolant activity presented in NUR.EG 0017 (Re f.11.1 2) .

Background information on critiu:.s production and fu4 operating entperience are eddressed in References 11.1 1 and 11.1-3, respectively.

@ 7-.Jarx r 11.1.1 Design Basis Radion:rivity in Systess and Ceeponents ig "

11.1.1.1 Eta ttt.C.colant Aetly.irl. h paraneters used in the calcula.

tien of the reactor coolant fission 7.oduct inventories, together with the pertinent information concerning the expected coolant clannup flow rate and demineralizer ef fectiveness, are smurized in Table 11.1 1. Calculated reactor coolant ractenuclide design concentrations are presented in Table 11.1 2. In these calculations the defective fuel rods are acsu:wJ to be present tt the initial core loading and to be unifo:uly distributed throughout the core; thus, the fission product escape rate coeffi.:lents are based upon average fuel temperature.

Tuel failure end burnup experience are presented in Chapter Is.

The fission product accivities in the teactor coolant durins operation witn sna11 cladding defects (fuel rods containing pinholes or fine cracks) are coeputed using the following differential equations:

Tor parent nuclides in tne cuolant:

dN c R,N, ,, ,

~*- -

1. + D, + h '"

Nc or daughter nuclides in the e.colant:

dN, R,N, .

' +

+ f,1,Ne- A,+D,+ '"

N,

. E DF, .

11.1 1 Revision 0 i

ATTACHMENT J

  1. ST HL AE-39o 6 PAGE J.L._ OF .) 7 unu:, unAr, (10/'h Y0 Y-where:

] N C

Concentration of nuclida in the reactor coolant (a ms/ gram)

N F

Populati n nuclide in the fuel (atoms) t - Operating time (seconds)

R -

Nuclide release coefficient (1/sec) - F F -

Traction of fuel rocis with defective cladding M

    • ?I ""* 58##**}

~ N 'I ******#

C A -

Nuclide decay constant (1/sec; A - l D -

Dilution coefficient by feed and blend (1/sec) - 5. A t DF 8, -

Initial boven concentration (ppm)

Boron concentration reduction rate (ppa /sec)

DT -

Nuclida decineralizer decontamination factor Q -

Purification or letdown mass flow rate (grams /sec)

S -

Nuclide volume control tank stripping fraction f -

Traction of parent nuclide decay events that result in the formation of the daughter nuclide subscript i refers to the parent nuclide.

Subscript j refers to the daughter nuclide.

The fission products are removed by decay, cleanup in the Chemical and Volu=e Control System (CVCS), and letdown to the Boron Recycle System (BRS). In the volume control tank (VCT), the fission products are assumed to be removed by decay only, with no degassing to the Caseous Vaste Processing Systen (CVPS).

The corrosten product activities, which are independent of fuel defect level, are based upon measurements at operating reactors. The corrosion product concentrationx are given in Tabis 11.1-2. These crud or corrosion product nuclidas are formed by activation cf eroded primary system materials. 3 Platecut and subsequent roerosion takes place throughout the primary system.

11.1.1,2 AVA 1 Cpntr C ank lgity112 Table 11.1 3 lises the ealcu-lated design and rea'.istic vapor activities in the VCT using the assumptions sum =arized in Tablec 11.1-1 and 11.1 6. The expected activities are based on a VCT purgt of 0.7 sefs.

11.1-2 Revisios, 0

.e -

d ATTACHMENT 3 ST HL-AE-3qd 6 PAGE JB 0F _2_2 INSERT "A" Source terms for extended burnup fuel are discussed in Section 11.1.6.

ATTACHMENT 3, P 9 lb.JAJ p tti F 1. ME 11.1.1.3 igeondary Coolant Attivity Model.

If defective staan generator (50) tubes exist, radionuclides from the primary coolant are introduced to the secondary system with the primary to secondary leakage. The resulting radionuclide concentrations in the secondary coolant depend upon the SC leakage rate, the naclide decay constant, and various parameters describing loss pathways in the secondary system.

Removal of radionuclides from the secondary system takes place by any of the following:

1. Condensate Dominera11:ar System treatment of condensr.te flow
2. Radioactive decay
3. Exhaust through the main condenser mechanical vacuua punp 4 ' Exhaust through the turbine gland seal
5. Main steaa leakage to the Turbine Generator Building (TCB) h
6. Condensate leakage to the sumps
7. Removal of nonrecyclable secondary samples
8. Removal thrugh the b1cvdown system.

The *model med to determine the concentration of nuclides in the SG liquid is given by a set of differential equations, as follows:

M, #" - 1.% BN,, M. A ,N,,

dt uhure:

e N,g -

Nuclide concentration in SC liquid, pei/5 N g Nuclide conventration in primary coolant. pCi/g N, -

SC liquid mass, g t -

Time, days L -

Primary to secondary leakage rate, g/ day a

.B -

Secondary system removal rate, g/ day A

g Radioactive decay constant, per day i

11.1-3 Revision 0 i

. . - -- . . = -

~

ATTACHMENT 3 s

' ST-HL.AE-390 6 PAGE f'[_. 0F A7-r

~~~iTPECS LTSAR 1

4gMM M*n W

Under the assumption of constant reactor coolant s.oncentrations, the nuclide concentration in the SG= liquid at equilibrium is given by:

. N,, = U" B + H, A, The nuclide concentration in the Reactor Coolant System (RCS) is assuced to be equal to the value. presented in Table 11,12 for design basis conditions.

Table 11.14 presents the parameters used to calculate the equilibrium secondary coolant concentrations.

Toc nuclides other than noble gases and trttium, these equations are used to cal:ulate the secondary coolant liquid conenntrations. We steam concentra-

.tions are then determined using the 53 partition factor applicable for the nuclide.

The model used to estimate secondary tritium assumes that the tritium becomes uniformly distributed throughout the secondary coolant and that the tritium concentration is at equilibrium.

In the case of noble gsses, the assumption made is that they are rapidiy tra oported from the water in the 50 and swept out of the vessel into the

. steam. Therefore, the concentration in the water is negligible, and the con:entration in the steam is equal to the ratio of the SG release rate to the ateam flev rate. These noble gases ars . removed from the system at the main condeaser.

The concentrations of nuclides in the secondary system vater and steam are given in Table 11.1-5, 11.1.2 Radioactivity Concentrations in the Fluid Systems Realistic Basis The parameters used to describe the South Texas' Project Electric Centcating Station (STPECS) reactor age given in Table 11.16 together with nominal values anc the range of values used by NURIG-0017 (Kef,11.12). The actual STPECS parameters have been used to adjust the standard coolant concentrations given in NL' REG-0017 (Re f.11.1-2) . D.e adjuatment of the standatd cociant

- concentrations vas performed using the standard formulae presented in NL' REG.

0017.

Specific acf.1vities in the primary coolant, SC vater, and the condenser, based upon the parameters of Table 11.16, are given in Table 11.17.

11.1.3 Tritiun production and Release to the Reactor Coolant There are:tvo principal contributors to tritium production within the STPECS systes: the ternary fission source and the dicsolved boron in the reactor coolant. Additional contributions ara made by' lithium-6, lithium 7, and deuterium in the reactor water and the B,C control rodt. Tritium production from different sources is shewn in Table 11.1-8.

Additional background information on tritium production is given in b Reference 11.1 1.

11.1-4 Revision 0 l

l .

l

) < ATTACHMENT 3 strEc5 ms.q ST-Hl.-AE-390 (.

PAGE is

- . - - - _.-.a OF_2_2,.n _ _

11,1.4 Activity in Radwasta Systems The design basta source terms for shielding and componenc f ailures for the Radvaste Systens are based uy,cn the concentrations shown in Tc.ble 11.1 2. The expected activities of the Radvaste Systess for effluent analysis tre based upon the concentrations shown in Table 11.1-7. The Liquid Ra:'veste System is further described in Section 11,2. the Caseous Radvaste System in Section 11.3, and the Solid Radvaste Systen in Section 11.4. The shimiding of these systems is described in Section 12.3.

11.1.5 Leakage Sources The systems containing radioactive liquids are potential sources of leakage to 9.e plant buildings and then to the envirornent. Lenkage from the primary system to the contairnent is expected to be less than 240 lb/ day. This 1sskate comes from such sources as valve packings. Leakage from the systems 1ccated in the Mechanical Auxiliary Building (MAL) is expected to be less than 160 lb/t.4y. This leaka.5e comes from such potential sources as put:p gland seals and valve packine.s. Total steam leaka6e in the TCB is expected to be less than 1,700 lb/ hour, as d'.scussed in Section 11.3.2.

These leakage sources and the resultin5 airborne concentrations are discussed core fully in Section 12.2.2.

Potential release points of radioactive ef fluents are discussed in Sections 11.2 and 11.3.

(- ,LO S CW T' 6' i

I 11.1-5 Revision 0 j

4 v

ATTACHMENT 5 ST HL-AE-3'fc 6 PAGE .J11 0F 37 -

INSERT "B" 11.1.6 The Impact of Ext.anded Burnup Fuel on Source Terms The source terms'presen'ed in Sections 11.1.1 through 11.1.5 are based on an equilibrium fuel cycle using discharge burnup of 33,000 MWD /MTU. The use of extended burnup fuel at STPEGS has been reviewed in NUREG/CR-5009, " Assessment. of the Use of Extended Burnup Fuel in Light Water Power Runctors" (References 11.1-4, 11.1-5) and has been deterinined to not significantly change the results previously presented in unfety analysis reports based on operation to 33,000 MWD /MTU discnarge burnup.

I l

t s

l l

I l

s 3

a

FiffACHMENT 2, STPtcs UrSAR 37.g(.AE- 3W' i 0F M-- J

[ PAGE _/2 BIf1RENCES f.td ion 11 1:

11.1 1 " Source Term Data for Vestinghouse Pressurized Vater Reactors",

VCAP 8253, Amendment 1. Vestinghouse Electric Corporation, July 1975.

11.1 2 U. 3. Nuclear Regulatory Com:nission, NUREG 0017 " Calculation of Releases of Radioactive Materials in Caseous and 1.iquid Effluents from Pressurized Vater Reactors (PVR Cale Code)",

April 1976.

11.1-3 " Operational Experiet.ce with Vestinghouse Cores". UCAP-8183 (1.arest edition).

n ,1 - 4 u . r . rJ w g& ;%-s, veryco-4 e M j w ,, i J rccs, "Arr<J t .3L h ]4 t~ P w < % 1,, c bdve tAV s oA 12.2 RADIATION SOURCES 12.2.1 Contained Sources The shielding design source terms are based upon the three genaral plant conditions of normal full povet operation, shutdown, and design bases events.

12.2.1.1 Ingrees for Normal Full Power Ooeration.

12.2.1.1,1 gesetor core: For the normal full power condition, a fission rate sufficient to sustain a power production of 4,100 We was used for the 14 foot high active core to determine the neutron source strength.

The radial source strength variation was based upon information provided in Chapter 4. The promet and short-lived fission gamma sources were also calculated for the 4,100 MWt core using data from Reference 12.2.1-1.

Irradiated incore detector and drive cable source strengths are used in determining shielding requirements and area radiation levels. These source strengths can be found in Table 12.2.13A.

12.2.1.1.2 Ceneral. Non Core Sources: The main sources of activity outside the primary shield complex during normal full power operation are N 16 from coolant activation processes, fission products from fuel clad defects, and corrosion and activation products. All shielding includes, as a design basis, the uaximum case of clad defects in fuel rods producing 1.0 percent of a core thermal power of 4,100 MWt (Section 12.3.2.2.2).

Each plant aystem is shielded according to the amount of activity present and the adjacent soning and access criteria. Ibse systems include the Reactor Coolant Systen (RC3), Chemical and Volume Cottrol System (CVCS), Boron Recycle System (BRS). Spent fuel Fool Cooling and Cleanup System (SPPCCS), and the Condensate Polishing Systen. Those sources that are contained in equipment of the Solid, Liquid, and Caseous Nte Processing Systems and the ECS Vacuu:n j Degassing System are described in Chapter 11. The following sections give the 1 activities or radiation source terns in the reactor coolant, in the auxiliary system proccas streams, and in the various plant components. .

12.2.1.1.3 Reactor ConlAn; The N 16 activity of the coolant in the primary loop compartments is the controlling radiation source in the design of the RCS secondary shielding and is given in Table 12.2.1 1 as a function of transport time in a reactor coolant loop (RCL). .The reactor coolant fission anti entrosion product activities are given in Table 11.12. $,

12.2.1.1.4 Sfur Generators _and Presmim: The steam generators  ;

(SCs) are located within the secondary shield in the Reactor Containment Building (RGB). The principal activity in the SG duriag operation is the N 16 activity given in Table 12.2.1 1. The deposited fission and corrosion product activity is important during shutdown and is discussed in Section 12.2.1.2.

The pressuricer is located in a shield cubicle just outside the secondary shield in the RCS, In addition, the pressurizer relief tank (PRT) is located just below the pressurizer. The radiation sources in the pressurizer and the PRT are tabulated in Tables 12.2.1-2A and 12.2.1-2B.

i 12.2-1 Revision 0 l

ATTACHMENT 3 ST HL AE Mob P M PAGE B 0F M l [y FSMw.

f STPgCS UFSAR q 12.2.1.1.5 Chemical and Volune centrol system: One of the functions of the CVCS is to provide continuous purification of the reactor coolant.

The major equipment items include regenerative and letdown heat exchangers (HXs), mixed be.d and cation bed dtainerslizers, reactor coolant filter, volume control tank (VCT), and charging pumps. D e Boron Thermal Regeneration (BTR)

Subsystem includes three BTR 10ts and the STR demineralizers. The Seal Water subsystem for the reactor coolant pumps (RCPs) includes the injection and return filters and the seal water HX.

Since the CVCS processes reactor coolant, the activity in the coolant going into the system (the letdown flow) is that given in Table 11.1-1. Table 12.2.14 gives a summation of this source by anergy group. The contribution to this source from the N 16 isotops is n;.t included because sufficient delay time is provided by the routing of the letdown line.

The radiation sources in the demineralizers, filters, VCT, and HXs of the CVCS arc given in Table 12.2.1-5. The source volume in each piece of equipment used in the shielding analysis is given in parentheses. A brief discussion of the location, function, and source of activity in each component is given in the fo11ovin5 Paragraphs.

All of the CVCS demineralit'rs and filters are clustered in one general area in the Mechanical Auxiliary Building (MAB) at the 41-ft level.

The mixed bed demineralizers retain the fission product activity, both cations and anions, and the corrosion product (crud) metals. The cation bed demineralizers are used internittently to remove lithium for pH control, and they supplement the mixed bed in rernving yttrium, cesium, molybdenum, and the crud metals.

The sp ific source strengths of the reactor coolant, seal water return, seal water injection filters, and letdown filters are included in Table 12.2.1-5.

The source for the reactor coolant filter correspnnds to r.n exposure rate of 500 R/hr contact, ne source for the remaining filters corresponds to an exposure rate of 100 R/hr contact. The filters were assumsd to be drained ot process fluid and vere emsidered to be honogentous sources.

The BTR beds are used to regulate the boron concentration in the reactor coolant. ney are utilized during load follow operations and in removing boron from the coclant as the nuclear fuel is depleted. These demineraliters also collect radioactive anions, such as iodine, which may have passed through the uixed bed. The specific source strengths for these beds can b6 found in Table 12.2.1-5.

The VCT is located in the MAB on the 41-f t level. The sources (Table 12.2.15) in the VCT correspond to a nominal operating level in the tank of 300 fts in the liquid phase and 300 ft 3in the vapor pFase.

The regenerative and excess letdown HXs are located in the RCB on El. 37 f t-3 in, and 52 ft, respectively. They provide the initial cooling for the reactor coolant letdown, and their sources (Table 12 2.1-5) include N-16 activity. The balance of the CVCS HXs are located in the MAB, where N 16 activity is not a significant factor. The letdown HX, located on the 10 f t ,

level of the MAB, provides second stage cooling for the reactor coolant prior l 12.2-2 Revision 0

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

ATTACHMENT e

ST HI-8.E 39 PAGE a 0F .XL. i k&/N'Nhh-

$7FECS 17p3AR

' to entering the domineralizers. The activity at this pett t is that coolant letdown, but it does not include the contribution trow the Wof16reactor isotops. Sufficient delay time is provided in the 1stdown line to make the H.16 source insignificent at the letdown hX. The spacific source strengths can be found in Table 12.2.1 5.

We BTa MXa include the moderating, letdown chiller, and letdown reheat units,  !

which 11 located on the 10.ft level of the MAB. The radiation sources  !

(Tab'- ...15) in this equipment are modified to accostne for activity remove.4 by t!o dominera11:ers upstream of the units. '

12.2.1.1.6 bnn htyrlt_1xalt.a: he radiation sources in the BR3 are l listed in Table 12.2.1 6 The source volume in each piece of squipment used  !

in the shielding analysis is given in parenthesen. The najor eqJipment items included in this systin are the recycle holdup tanks (RHTs) and the recycle evaporator with its as.*ociated equipment, i.e., feed deafneralizers and filter, condensate desirer =11rer and litter, and concentrates filter.

Radiation sources in the variotus puapr. are assumed to be identical to the ligold sou w s in the tank from which the pump takes suction.

All of the BRS deninera11rers at d (11 tera are clustered in one general area on the 41.ft level of the MAB.

The evaporator feed demineralizers are located upstream of the holdup tanu and contain mixed bed resins which remove nongaseous activity from the reactor coolant directed to the holdup tanks.

The avaporator cer.densate dermineralizer la charged with anion resin to remove any boron and iodine activity which may be carried over with the evaporator condensate.

We RHTs, located or the 10.fr. level of the MAB, are each equipped with a diaphrap- Cases vWh flash from the react 0r coolant letdovn to the holdup i tanks are retained - @ : the diaphragm, Periodically, the gases are vented to i the caseous Waste Processing System (CW?S). D e radiation sources in the holdup tanks are based uien 50 percent of the gaseou.s activity flashing into the vapor phase.

%e recycle evaporator feed filtsr and condenaste filter are located downstrema of their respectiva demineralizers, and serve to retain particulates and any resin finos which may ascope from the desineralizers.

The maximum radiation sources on these filters are given in Table 12.2.16.

The sources for the feed filter correspond to a radiation level of 100 R/hr c u ntact.- Yne condensate filter sources result in 1svela of less than 1 R/hr contact. The resultant radiation sources on the concentrates filter correspend to an exposure rate of approximately 10 R/hr contact.

12 . 2 .1.1. '/ Seent N1_1; col CopUne and Clf anue SyLifJn: The radiation sources in the SFPCCS are given in Table 12.2.17. The STPCCS demineralistra and filterr. are located at Z1. $1 f t 6 An. of the MAB. The demineraliters and flisers are used to maintain water clarity and remove activity released during refueling operat17ns and the subsequent fuel cooling period. The filter sources correspond to an exposura rate of 100 R/hr at contact. The STPCCS ,

17.2 3 Eevision 0 Lv w , < , , , --.e , . . .na w . _ _ . _ . . _ _ . _ _ _ _ _ _ . . _ _ _ _ . _ . . _ _ _ - _ _ - _ _ _ ,

g ,;

~4TT4CHMENT 3 6~

3T HL AE J'14 PAG E .1L_._ 0F _.*l 7 . _ d*-f' strzCs ursAR Q  ;

pumps and MXs are located at E1. 21 f t.11 in., and 42 f t.6 in., respectively. l of she Tuel Handling gutiding.

12.2.1.1.8 Cedinsate polishina svateo The design bases adiation sources for the Condensate Polishing System are given in Table 17 2.18. The sources are based on the renoval of radioactive contaminants by s.ie '

s'esineraliser beds assuming 1 gal / min primary.to. secondary 50 leakage 1 percent fuel clad defects, and a danineraliser run time of 42 days for the mixed beds and three days for the cation beds.

4 The desineralizers and cation and anion regeneration tanks are located at E1.

29.ft of the Turbine cenerator Au11 ding (705). here are three high total dissolved solids (TDS) tanks, one for the cation. bed desineralizers and two for the mixed. bed denineralizers. Each tank is assumed to contain one batch of regeneration solution. Ninety percent cf the activity in the regeneration tanks to assumed to pass to the high TDS tanks with the rest going to the lov TDS tank. The high TDS tanks are in the yard near the TC5.  ;

. 12.2.1.2 Sources for Ebu1ARyn conditions. The core ganaa sources (Q (af ter shutdown) are used to establish radiation shielding requirements during

% s refueling operatio nd during shipment of spent fuel. ne bases for the i core average sou Us engths is a three rer; ion equilibt'tum cycle core at end.

y of. life (Sesti 3 4.2) These source strengths, per unit volume of homogenized 4 core, tabul ted Table 12.2.1 38 for various times after shutdown.

7p deposited corrosion product activities as a function of time for the RCS 4 .

/ re given in Table 12.2.19.

An additional system requiring shielding is the Residual Heat Removal (RHR)

.' System. The specific source strengths in the RHR loop are given in Table 12.2.1 10 The RHR loop is placed in operation after primary system pressure N and terparature have been reduced to 400 psig and 350'T. The sources are saximum values with credit taken for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of activity decay and purification.

12.2.1.3 Sourcu for Desien nases E.'ents. The sources for the Design Basis Accidsnt (DBA) used to detaraine the 30. day exposures to control room personnel, the f. hour exclusion zone boundary dese, and low population zone doses are presented in Sections 6.4 and 15.6.5, respectively.

, 12.2.1.4 11eht '..un Proc e n s Fleina. All piping containing radioactive fluids is shown on piping diagrams and piping composites. Routing of all piping shown on the composite drawings was performed and checked by engineering prior to installation.

12.2.2 Airborne Radioactive Material Sources This section deals vitir the models, parameters, and sources required to evaluate airborne concentrations of radioraclides during plant operations in various plant radiation areas where personnel occupancy is expected.

1.aakage sources are dependent upon the concentrations of radioneclides in the primary systen, secondary system, spent fuel pool, and the refuelin pool.

The. assumptions and parametes s required to evaluate the isotopic airLorne concent' rations in the various applicable regions are listed in table 12.2,21.

12.2 4 Revision O'

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

ATTACHMENT 3 ST HL AE Y10 6 PAGE A L__ 0F II,,I N[M fy  !

STPECS UFSAR f*g 4 D,7 The CVCS and the SFFCCS were designed to purify reactor coolant through ion i exchanprs after reactor shutdown and cooldown. This ensures that the effect  ;

of activity spikes will not significantly contribute to the Containment airborne activity during refueling operations. The contribution to airborne ,

activity due to reactor vessei head renoval is considered negligible as the  ;

RCS Vacuus Degassing Systen (Section 11.3) will remove this activity prior to head removal. The detailed listing of the expected airbome isotopic toncentrations in typical accessible regions is presented in Table 12.2.2 2.

The final de cign of the plant ensures that the expected airborne trotopic concentrations in the typical accessible regions are below the maximum permissible concentration for the critical organ f or the appropriate isotope for occupational workers, as adjusted on the basis of expected occupancy in ,

the regions (i.e., access to containment is expected only 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per week during normal operation). ,

12.2.2.1 t[g4tLfor Calculatina Airbornt Concentrations.

Plant areas with airborno radioactivity are character 1 red by a constant leak rate of a radioactive source at a constant source strength with a constant exhaust rate el the contaninant, This leads to a peak or equilibrium airborne concent';ation of the radioisotope in the regions as calculated by the

, followies equation:

C (t) - (IA), A, (PF), (i.eAi') (Eq. 12.2.2 1)

(VA,,)

where:

(1.R), - leak c,r evaporation rate of the ith radioisotope in g/sec, in the applicable region, and A, a activity concentration of the ith leaking or evaporating radioisotope in pC1/g (PF), - Partition factor or the fraction of the leaking activity that is airbome for the ith radioisotope A, - total removal rate constant for the ith radioisotope in sec*'

from the applicable region

- ( A,i + A.)

(A , and A, are the removal rate constants in see due to radioactive decay and the exhaust from the applicable region respectively for the ith radioisotope) t - time interval between the start of the leak and the time at whica the concentration is evaluated in seconds V - free volume of the region in which the leak occurs in em' 12.2 5 Revision 0

'IIihCliMENT 3 S1 til AEJct 0 6 PAGE JJ _0F _? 7 STPECS UFSAR =

C.(t) - airborne concentration of the ith radioisotope at time t in

.vCi/ca' in the applicable region Trom the above equation, it is evident that the peak or eqAlibr!'ta concentratton, C , of the Ith radioisotope in the applicable rtglon vill be given by '.he following expression:

C,.., - (1.R) A, (PT), / (V A,,) (Fq 12,2.2 2)

Vith h'.gh exhaust cates, this peak concentration vill be reached within a few hours.

A (MSc9 T C 1

I l

12.2 6 Revision 0

. ATTACHMENT 3 SbHl. AE 3'(o 6

- PAGE_:LL._O_F 2 7 INSERT "C" 12.2.3 The Impact of Extended Burnup Fuel on Sourco Terms  !

The source terms presented in Sections 12.2.1 and 12.2.2 are based on an equilibrium fuo) cycle using discharge burnup of a3,000 MWD /MTU. The uno of extended burnup fuel at STPEGS has boon reviewed in NUREG/CR-5009, "

Assessment of the Use of Extended Burnup ruol in Light Water PoWor Reactors" (References 11.1-4, 11.1-5) and has boon determined to not significantly change the results previously presented in safety analysis t reports based on operation to 33,000 MWD /MTU dischargo burnup.

L t

4,, y' s ,- .- - _ , - a.-- . - . , y .w. ,--.v.,*-y- -wy -- - -~~ + -- ---weo~7-e-'~t==--**'"*-'F***---**F-"

. .. ATTACHMENT 3 s . ST HL. AE.3'W le K I,_ $ . /, 0 PAGE 25__0F 33 =

gyQM

Trsc5 vrsAR tg -

APPENDIX 15.A r38110N F10DVCT INVENTOR!LS 1$.A.1 Activities in the Core The calculation of the sore iodine and noble gas fission product inventories is performad with the OF.!CgN code (Refs.15.A.1 and 15.A.2). All inventories are based upon a core power level of 4.100 Wwt. The core activities are given in Table 15.A.1.

i 15.A.2 Activities in the Tuel Pellet. Cladding Cap The fuel. clad gap activities are determined using the model stven in Regulatory cuide 1.25. Thus, the amount of activity accumulated in the fuel.

clad gap is assumed to be 10 percent of the iodine and 10 percent of the noble gases accumulated at the end of core life, except Kr.85,1 127, and 1 129, where it is assumed to be 30 percent of the core activity. The gap activities are given in Table 15.A 1.

15.A.3 Concentrations in the Coolants The concentrations of various radionuelides in the prisery and secondary coolants tre presented in section 11.1. The concentrations of iodines and 4 noble gases are , en in Table 15.A.2. Section 11.1 presents the assumptions i by which these < ..centrations were calculated. '

Daring full. power operation, thi primary coolant concentration Technical Specification limit for continued operation up to 48 houra is 60 pC1/g of  !

equivalent 1 131. The iodine licit takes into account an iodine spike l occurring because of a previous power transient. The reactor coolant  ;

concentrations corresponding to these Technical Specification limits are given in Table 15.A 4 The secondary coolant Technical Specification limit is 0.10 pC1/g of equivalent 1 131. The concentrations corresponding to this limit are given in i Table 15.A.S.

The effects of todine spiking caused by an accident have been acesunted for by increasing the iodine source ters in the primary system upon depressuritation or reactor shutdown. The sodine concentration during the iodine spike is governed by the following differential equation:

AC - E _ AC dt H where:

C - todine nuclide concentration, pC1/g F, iodine release rate from fuel, pC1/aec 15.A.1 Revision 0

, ATTACHtFNT3 St Hl. AE-3'M G sTrr.cs Ur$AR PAGE 1(e._0F _21_.

N - aass of primary coolant, g .

A - radiological decay constant, per see The release rate from the fuel has been increased by a f actor of 500 (over the equilibriu:s condition release rate) to model the effect of the spike. The iodine appearance rates in the reactor coolant for normal steady. state creration at 1 pCi/g of does equivalent 1131 and for an assumed accident.

initiated iodine spike are given in Table 15.A.6. The iodine appearance rates for the sGTR are given in Table 15.A.7.

rogar v 15.A 2 Revision 1

AT TACHMENT 3 ST HL AE $1 %

, PAGE A 2_.OF_1 2_

INSERT "D" 15.A.4 The Impact of Extended Durnup Puel on Source Torms The source terms presented in Sections 15. A.1 through 15. A.3 are based on an equilibribe fuel cycle using discharge burnup of 33,000 MWD /MTU. The use of extended burnup fuel at STPEGS has been reviewed in NUREG/CR-5009, " Assessment of the Use of Extended Durnup Puel in Light Water Power Reactors" (References 11.1-4, 11.1-5) and has been determined to not significantly change the results previously presented ih safety analysis j reports based on operation to 33,000 MWD /MTU discharge burnup.  ;

The increase of fuel burnup will not significantly impact the l radiological consequences of both LOCA and non LcCA accidents discussed in Chapter 15 or the control room operator doses presented in Section 6.4. NUREG/CR-5009 predicts a possible 20%

increase in the offsite thyroid dose as a result of a Puol Handling Accident due to an increaso in tre release fraction Thl, of I-131 into the fuel-clad gap for extended burnup fuel.

increase has boon previously reviewed for STPEGS and found to be acceptable (Reference 11.1-5).

i 1

Retrieved from "https://kanterella.com/w/index.php?title=ML20079L225&oldid=2758743"