ML20148G286

Radioactive Liquid Effluent Sys Evaluation Using Program Liquid Effluent Based Operating Guides Rev II

ML20148G286
Person / Time
Site:	Rancho Seco
Issue date:	01/31/1988
From:	Martin D, Murphy P, Story H SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
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ML20148G263	List: ML20148G258 ML20148G286
References
TAC-56033, NUDOCS 8801260461
Download: ML20148G286 (32)

Text

.

t RADIOACTIVE LIQUID D7IIENT SYSTDE EVAUJATICH USING PROGRAM LEBOG (Liquid Effluent Based Operating Guides)

REVISICH II Prepared by the DNIRCtMEL PROMYICH DEPARDE?T at the RANOO SECO NUCEAR GDERATING STATICH Herald, California January, 1988 written by:

i MA-W l

Peter Murphy, HeQth %ysicist i

{

reviewed by:

he '

Harvey Stqff, b @ isc{7 Health Ihysics & Chemis\\ ry Services l

w roved by:

(

.pr 'Ibnald T-Martin, W Dvilxtenut.al PrcractierY kha khgKO500031g 1 880113 1

P PDR L

TABIE OF CONITNIS BEIS I.

INIRXXJCI' ION 1

II.

OVERVIEW (EXECUI'IVE SUM 4ARY) 1 III. METHODS

a. Failed Fuel Defect 2

b. Primary Coolant Activity 6

c. ~iquilibrium Concentration in Secondary 6

d. Activity Directed to RHLTIs 8

e. Slulosable Demin Operation 10

f. Dose Calculation 11 IV.

RESUIIIS la V.

CDNCIIJSIONS 12 VI.

APPENDICES A: References A-1 B: Code Results B-1 C: M of IEBOG C-1 D: Ncnenclature D-1 E: Verification E-1 l

F: Sumary of Revisions F-1 1

s 1

Eidx mrIm h Technical Specifications for primary system activity and primary to secondary leakage at the Rancho Seco Nuclear Generating Station were written to satisfy ~ 10 CFR Part 20 for the air dose resulti19 fran secondary side safety relief following a steam generator tube leak.

h original plant design gave little consideration to the pcesibility that operation with small primary to secondary leakage below the Tehniual Specifications oculd jeopardize the plant's standity with respect to 10 CFR 50 Appendix I.

These numerical AIARA objectives restrict the whole bocq dose resulting frm the plant liquid effluent to 3 mrem per year.

Plant experience has shown tha'.,

given the extremely small dilution available frun Clay Creek, the guidance presented in 10 CFR 50 Appendix I is considerably nore limiting than Part 20.

Ib gain a better unckrstaniing of the plant's twe creratirg limits, the carpIter code IEBOG (Ic. quid Effluent Rwd Operatirg GJir es) was developed to i

model the effluent related plant systems and predict the anaac frun plant operation over routine ranges of failed fuel fraction and primary to secondary laak ratec.

This r M = ant presents the remilts of tnat evaluation and describes the assunptions and methods used within the code.

CODE OVERVIEW Pregam IEBOG is a simple but conservative ocmputer code written in the IIH basic language.

It calculates the dose frm liquid effluents at Rancho Seco over a calender year.

Historically, the adult whole body dose has been the limiting case, and two radi cesiuta isotopes (Cesium-137 and Cesium-134) have contributed roughly 97 percent of that dose (see Reference 12)

The code uses average plant operating parameters to model the transport of cesium frm the reactor coolant thruxfi the Once 7hrocgh 5 team Generator (OISG) leakage to i

the plant affluent streams.

The calculational prmadure in IEBOG ocntains seven steps:

(1) h data section defines, the pl r.t ;pecific system parameters and dose related inforretion for the isotopes of o;ncern (2)

Tne specific activity of Iodine-131 in the prinury coolant is cal.culated for a given failed fuel fraction using t'se relation derived in a l

Babocck and Wilcox Igort (RefcTence 1).

Tne cuv. aid. ration of Cesium-137 is l

then obtained frm an avar$9e ratio of Cs-137 to I-131 u.=reation, drawn frun Rancho Seco historicC data.

(3) The ocde dotermines the aquilbrium cu x:ad. ration of Cesium-137 in the secordary systet frun a alffereri~ial equation dammibLg the relative rates of activity incu arrt outflow, nc best ergineering estimate of the secondary side ;eakage, the total activity that Jeaves the turbine plant to the Regenerative Holdup Tanks (RHLTrs) is calculated.

(4)

The amount of activity to be discharged may be r*W=1 further by the l

l operatien of sluiceable deminenlizer skids which may be lined up between l

RHL7Ps to prmaaa the effluent pric to release to the Retention Basins.

l 4

2 (5)

'Ihe amount of Cesium-134 has traditionally been 55 percent of the Cesium-137 scurce tem.

Deteminim the amount of Cesium-134 to be relmaad frm this ratio, the offsite dose is calculated with the hand method given in the Offsite Dose Calculation Manual (ODm) Revision 2.

(6)

In addition, the release source tem accounts for transfers of slightly cmtaminated water frm the primary side (Demineralized Rea< tor Coolant Storage Tank or IRCST) to the RHUr for release into the plant effluent.

(7) h doses calculated for various ocabinations of failed fuel fracticn and primary to secondary leak rate were nomalized against the Appeniix I value for a calender year e===4rg a 61 percent availability.

LEBOG was verified by a hand calculation for the case of 0.05% failed f'el and a 0.10 gpn primary to secordary leak rate. 'Ihis represented approximately the status of Rancho Seco prior to its forced shutdown in W2 of 1985 except that the actual leak rate varied between 0.04 and 0.07 gpn.

IEBOG results indicate that plant cculd operate at leakage up 0.2 gpn with the fai'.ed fuel present at shutdown.

ME'110DS

'Ihis section hibes all of the equations ard assumptions used in IEBOG.

For convenience, Appendix D lists and defines the variables (with units) found in the code text.

a. Failed Fuel Defect

'Ihe following diagram (taken frm Reference 1) h ibes the production of I-131 in the fuel and its migration into the coolant.

\\

/

\\ PL{NVM

'UEL cu. c COOLM T p

V

" 2. _

r V

=N

, 3 y

g e

'Ihe variable N stands for the nurrber of I-131 atoms in the region specified by the subscript.

'Ihe letters f, g, c represen'. the fuel, plenum gap, and the coolant respectively, h rate of I-131 ation in the fuel equals the product of the thermal power P (N), the tL A yield Y of I-131 (fraction), ard a constant relating the number of nuclem events to the energy produced (in fissions /W-sec):

(dN /dt) prod (atczns/sec)= P(W)*Y*F (3.3E16 atcras/W-sec)

(1) f l

l s

3 Note that the nu:rber of fissions is the number of ctoms changed and may therefore be interchanged with the dimension "atcras" dir~:tly.

h migration of I-131 into the plenum gap is modelled as e diffusive process where the rate is proportional (by a constant factor vi) to h nmber of atoms present in the fuel:

(dNg/dt) escape (atcras/sec) = vi (1/sec)

• Ng(atcras)

(2) h derivation of the escape coefficient vi is dimtM in part b of this section.

The rate of radioactive decay can be calculated as the product of the arter of atctns and Cm iecay constant, i efined as r

'dNg/dt) decay = r

• Ng (3) h net rate ot' change for I-131 in the fuel is then the sum of the production, escape acd decay terms:

dN /dt = F*P*Y - r*Ng - vi*Ng f

= F*P*Y - (r + vi)

• Ng (4)

Since v1 is three orders of magnitude smaller than r, that sum may be replaced by r so that dN /dt = F*P*Y - r*Ng (3) f Similarly, the equation for the not rate of change for I-131 in the plenum gap is:

dN /dt = vi*Ng - r*Ng - v2*Ng g

l

= v *Ng - (r + v2)* Ng (6) l where the first term is the influx of activity from the fuel, the second is the loss by radioactive decay, and the last describes the escape of I-131 into the coolant.

As a matter of conservatism, the escape coefficient v2 is set at unity meaning that 100 percent of the gap activity passes into the coolant every second.

Since the value of r is five orders of macnitude smiler than v2,itten:it may be neglected in the sum which is then unity. 7m equation my be rewr i

dNg/dt = vi*Ng - Ng (7)

I h overall rate equation describing the coolant activity incln h the rate of influx frcan the coolant, loss through radioactive decay and losses through the

4 i

i Makeup and Purification systen. 'Ibe prtrbction tea is defined (dN /dt) prod " V *D*Ng (0) o 2

where D represents the fraction of failed fuel.

Note that this is inconsistent with the balance performed for the gap.

While not algebraically correct, this assunption is reasonable given the accuracy of the derivation.

Note that not including the defect D in the gap balance would cancel out this factor in the subsequent analysis.

h cleanup term can be written as (9)

(dN /dt)denin = k

• Nc c

where the cleanup coefficient is:

B (gal /sec) * (1 - (1/DF))

k (1/sec) =

(10)

V (gal) and where B is the average pr-a flow rate and V is the systan volume.

DF stands for the decontaminatim factor of the purification systen:

concentration of letdown DF =

(11) concentration of makeup h equation for the net rate of change in the coolant is then; dN /dt = v2*D*Ng - r*Nc - k*Nc e

v2*D*Ng - (r + k) *N

=

c D*Ng - (r +k)*Ne (12)

=

again using unity as the value for v2' When the plant reaches equilibrium, the rate of production is matched by the losses so the net charxJe is zero.

'Ihe three balance equaticos (5,7, ard 12) becane:

O= P*Y*F - r*Nf (13)

O= vi*Ng - Ng (14)

O= D*Ng - (r + k)*Nc (15) l Solving each equation for Ng,

'.4, and Nc respectively yields:

l g

1 P*Y*F Ng =

(16) r

5 1.

Ng=

s'y*Nf (17)

Ne =

(18)

(r + k)

Substituting equation 16 into equation 17 and then equation 17 into equation 18 results in:

D*vi*P*Y*F Ne =

(19) r* (r + k)

Independently we know that the number of I-131 atms in the primary coolant is defined by its radioactivity:

r(1/ soc)* Nc (atms) 37000 (atcms changed /uci)

'Ibe total activity in the prianary system is also A(uCi) =

C (uCi/ml)

• Vres (ml)

(21) where C is the specific activity of I-131 in the primary system which has a volume V in ml.

'Iberefore, we can eliminate the activity A by substituting equation 21 into equation 20 C*Vres=r*Nc / 37000 (22) and solve for the number of atms in the coolant Nc = (C

• Vrcs

• 37000) / r (23)

Substituting this into the balance equation 19 derived earlier yields:

C*Vres *37000 D*Vi

• P *Y

• F (24)

=

r r * (r + k)

Cancelling the r in the denominators arxi solving for the wmhation we arrive at D

• v1

• P *Y

• F Vrcs

• 37000 * (r + k) l

'Ihis can of course be nultiplied by 1E-6 to give the oormhation in units of Ci/ml.

6 b.

Primary Coolant Activity USAR Appeniix 14D uses a value of 2E-8 (1/sec) for the fuel escape coefficient under accident conditions and a value of 1.3E-8 (1/sec) for both design and expected cor.ditions. A principle assunption in the IEBOG analysis is that the fuel defect is linearly proportional to the concentration of I-131 in the primary coolant.

'1he cc/c.stant of proportionality was that inferred in the USAR, and the vi coefficient was backcalculated to insure that the I-131 concentration predicted by IEBOG for a c3 ven fuel defect was consistent with i

the USAR.

Although the USAR indicates that the concentration of Cs-137 in the primary coolant is larger than that for the I-131, historical data on the primary coolant activity indicate that the equilibrium amount of Cs-137 normally runs an order of magnitude lower than the I-131.

Chemistry data frca the last year of operation (1985) are not representative of steady state operation because the plant had only a 30 percent availability and experienced numerous startups ard shutdowns.

'Ihe previcus year (1984) provides a more realistic estimation of the ratio between the two isotopes 'nticipated in 1988. Averaging the Cs-137 to I-131 ratio fran the daily sanple cnalysis gives a result of 0.14.

While the ratio predicted by the USAR differed from our cperational history, the magnitude of the I-131 ccr =u'anticn was cmsistent.

'lhe primary side I-131 ocnoentration couwspcialing to the 0.05 percent failed fuel (Reference 14) shows excellent agreement with the 0.1 uCi/ml observed at shutdown (Reference 16).

In IEBOG the cr.noentration of Cs-137 in the primary coolant was determined by nultiplying the I-131 concentration by the cesium to iodine ratio.

'Itr.n the rate, defined as L,

at which Cs-137 enters the secondary side can be calculated for a given primary to secondary leakage W with the equaticn:

L(Ci/ min)= W(gal / min) *3785(ml/ gal) *C(Ci/ml;

• AVAIL (26)

'Ihe variable AVAIL here means the plant availability exprecsed as a fracticn; its inclusion accounts for the redt.ction in source term by -tion or leakage when the plant shuts down.

c. Equilibrium Coruuiuation in Secordry Considering the seccndary systan as a control volume, the rate at which activity changes can be described by a differential equation containing a production term (equation 26) frun the inleakage of primary coolant and loss terns associated with operation of the Condensato Polishing Denineralizers ard secondary side leakage.

Specifically, the rate of removal by the Condensate Fchshing Demineralizers can be calculated by:

R(C1/ min) = K (1/ min)

• A (C1)

(27) where A is the total amount of Cs-137 activity in the secondary, and K is the

7 i

remwal rate coefficient which can be calculatai utrough equation 10.

'Ibe activity loss frm % leakage is:

H(Ci/ min) = C (C1/ gal)

• Z (gal / min)

(28)

In this case, C is the concentration of the e.amndary side and Z is the s3condary leakage rate.

'Ihe time rate of change of secondary side activity is dA/dt =

L R

H C*Z (29)

K*A L

=

Since the secondary side concentration is C=

A/V (30) where V, the secondary volume, is a constant.

'Ibe time rate of charge of Cs-137 concentration can be calculated using equation 29 as:

dC/dt = (1/V)

• dA/dt

= (1/V) * (L - K*A - C*Z)

(31)

Solving equation 30 for A aM substituting that into equation 31 yields dC/dt = (1/V) * (L - K*(C*V) -C*Z)

= (1/V) * (L - C*(K*V + Z))

= (I/V) - C*(K + (Z/V))

(32)

We can rewrite this as 1 - k *C)

• dt (33) dc = (k 2

whare ki aM k2 are cmstants.

For convenience, define a dumy velable u=ki - k *C 2

-(1/k )*u +(k /k )

(34) then C=

2 i 2 ard dc = -(1/k ) *du (35) 2 Subst'.tuting equations 34 and 35 into equation 33 :wts:

-(1/k )

• du = u

• dt (36) 2 Separate variables:

du/u

-k2

• dt (37)

=

Integrate both sides:

In(u) = -k2*t

+ CONST1 (38)

8 where OCET1 is the constant of integration. Raise both sides of the equation as a power of e, the mtural logarithm exp (-k *t + CONSTl) u =

2

= exp (-k *t)

• exp(CONSTl)

(39) u 2

Since tu second factor is essentially a constant, define a new constant:

u = CONSr2

• exp (-k *t)

(40) 2 Resubstitutiry the original variables gives:

i - k *C = CONST2

• exp (-k *t)

(41) k 2

2 At initial cxxxiitiws (plant startup), t equals zero so the exponential term is unity. We also know the concentration C is zero. 'Iherefore, the value of the constant equals:

CONST2 = ki (42) s M tuting equation 42 into eq ution 41 yields i - k *C = ki

• exp (-k *t) k 2

2

-k *C = -ki+ki

• exp (-k *t) then 2

2 k *C = ki * (1 - exp(-k *t))

(43) 2 2

'1he general solutico describirq the time dependence of the secondary side Cs-137 concentration is then C = (k /k ) * (1 - exp(-k *t))

(44) i 2 2

As timD t goes to infinity, the value of the exponential approaches zero, ard the equilibrium concentration is then (I/V) k /k2 (45)

C=

=

1 (K + Z/V) where ki ard k2 were defined inplicitly in equations 32 ard 33.

Because Cesium-137 has a lcog half-life (approximately 30 years), no credit is taken for radicactive decay.

d.

Activity Directed to RHUrs

'Ibe total activity reaching the PHUrs comes frm essentially two sources. 'Ihe first is leakage from the secondary side which collects in the turbine building grade level sunps and is then punped to the RHUrs.

'Ibe second is water frm the IRCST transferred to the RHUrs for release offsite when the production of slightly contaminated water frm the liquid radwaste systans aw=@ the site demani ard storage capacity.

9

'Ihe leakage source term can be calculated by the equation Q1 (Ci) = c (C1/ gal)

• Z (gal / min)

• T (uda)

(46) where, in the IEEOG analysis, the time T is a cale.xler year adjusted for plant availability; that is, T(min) = 365 days

• 24(hr/ day)

• 60(mirVhr)

• AVAIL (4*i)

'Ihe average secondary leakage rate was estimated to be 50 gpm using two different methcxis.

One method involves gauging the secondary system makeup, and the other requires estimating the total systan leakage.

Ma)mup to the secondary plant originates with the Denineralized Makeup Water

([M1) systen, a fraction of which is directed to the Condensate Storage Tank (CST). In turn, a fraction of that water enters the secondary as makeup while ths rest supplies other plant needs for denineralized water.

'Ihe operations staff estimate that in any given 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of plant operatic 1 at full power, the DW system is lined up to the CST for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In the last six years of operation (1980-1085), the largest amount of water prmaaaad per anrum through DM was 44 million gallons. At a 61 percent availability, this translates to roughly 140 gpn. Assuming that a fourth of that water goes to the CSr and all the Csr water was used solely for secundary loop makeup, the seemidary leakage would be approxhnately 35 gpn. A 50 gpm leakage would then be conservative.

A meuber of the ruclear engineering staff has surveyed the secondary sidis leakage to evaluate the possibility of rerouting polished condensate (seal water) back to the weersier instead of directing it to the sunps.

He estimated the volume flow rates of the total known sources of water at a 20-35 gpn contirmous loss.

If 10-20 gpn is allowed for other sources of leakage, the total would ccane to a mavi== of 55 gpn.

Since this estimate does not include evaporation Irwaaa and recycliry of turbine buildirq sung water back to the Miscellaneous Liquid Radwaste systan, an estimate of 50 gpn is a reasonable approximation.

Given the consistency of the two methods, this number was ueed as a "best engineering estimate" of the secorriary inmaan.

'Ibe second source of activity to the IHJrs is the transfer of water frun the primary side. When the plant operates with prinary to secondary leakage, the fraction of turbine building sunp water returned to the liquid radwaste systan incraaana.

If the slightly contaminated water resultirq frun this prma==

evaada the site storage capacity, it nust be relaaaad offsite in liquid effluent to prevent tank overflows.

Historically (Reference 15), the worst year for these releases was 1985 in which 787,000 gals of water (average concentration of 1E-7 uCi/ml of Cs-137 and a proportionate anount of Cs-134) were directed to the RHUrs. 'Ihe IEEUG code assumes that 1 million gallons of water at those wrishations will be relaaaad wjthout prmaaaing by the sluice Demin Skids.

'Ihe amount of tritium relanaad in liquid effluent fran all plant sources normally varies between 60 and 90 Curies per year.

'Ibe average over the last six years (excludirq the abnormally high value in 1984) is 75 Curies. IIBOG assumes a value of 100 curies. Given the chemical nature of tritium, no credit is taken for renoval by demineralizatical.

1

'Ihe source term frun the primary is then:

1 10 Q2 (C1) = Vdrest (gal)

• 3785 (ml/ gal)

• 1E-13 (Ci/ml)

(48) j and the total activity in the IM.Trs is Q (C1) = Q1 (C1) + Q2 (C1)

(49) e.

Sluiceable Domin Operation Recent plant modifications allow for prmaaning of RHUr water through sluiceable demineralizer skids.

The equipnent will be operated in a htch mode and is sized to permit at least one pass through the skids for each RHUT.

Raamd on historical data, the annual amount of water passing through the A and B IMJIs can be estimated at 45 million gallons for a 61 per nt availability.

This equates to rurply 86 gpn. The demin skids have two trains sach with a rated flow of 50 gpn, and both may be operated in parallel for a total capacity of 100 gpn.

The original design wWt had the B RII7f acting continuously as a receiver tank.

Its contents would be pW through the desin skids to the A RIIII for discharge to the retentim basins. With batch operation, the activity on the downstream side of the hins is sinply (Q) downstream = (1/DF)* (Q) upstream (50)

Tb account for the fact that the instantaneous rates of waste water flow to the R! itis may exceed the capacity of the skids for short times and further that this equipnent will sanetimer be out of service, it is cx:nservatively assumed that only a fraction of the waste can be prnw this way.

The resulting scurm term of Cs-137 is then:

Qeff (C1) = Q * (1-f) + f* Q/ DF (51)

In their contract bid the manufacturer of the sluiceable damins, Pacific Nuclear, estimated a decontaminaticn factor of 1000 given the resins and the anticipated inlet wwsd. rations and water chenistry. However, this figure is based on engineering jt%---Ot, and other guidance (see References 8 and 9) suggest that this may be overly optimistic.

As a matter of ocnservatism, a value of 50 was used.

Actually tlne sluice damins can be cperated in two main ocnfigurations.

The first, the "straight-through" mode hibed above, is the most effic.'.ent. It does, howevar, have the diandvantage in that, with B RHUr acting contirmously as a receiving tank, the chemical quality of the inlet water cannot be easily controlled.

Bad chemistry may ruin or strip the resins. Cbnsequently, the plant would either run in this ocnfiguration and risk losing equipnent capacity in the event of a chemistry problem or suffer a systemic reduction in capacity by using only one tcain arx! keeping the other in reserve.

The alternate operational mode would be to isolate an R117f when full and recira. late through the skids.

That is, the skid booster punp wenid move water frun the RHUr through the sluice demin skids and retun it to the same RHUr.

Recent analyses suggest that the effectiveness for this secxand mcde could match the first with relatively small increases in processing time.

11

f. Dose Calculation Historical data show that the average relative abundance of Cesium-134 to Cesium-137 in liquid effluent is 55 percent.

Since the two isotopes are identical chemically, operation of the sluice demins will not affect that ratio. 'Iherefore the annInt of Cesium-134 available for release is:

Ocs134 (C1) =.55

• Qeffes137 (C1)

(52)

'Ibe resulting offsite dose can be calculated using the ham method described in the ODCM. 'Ibe fon:ula is:

3 Qi (C1)

• A j (mrem-ft /Ci-sec) i D(mren) =

(53) 3 F (ft /sec) where Qi is the total amount of activity of isotope i released during the time period of interest, F is the total plant effluent flow rate which cm be set by operaticnal precedure to have a mininum of 8500 gpu (18.93 cfs)

A q is a cxzbitwa dispersion and dose factor specific to each and i1sotope i and organ j (whole body).

'Ihe sum of the h fran the two oesiums is a=W conservatively to contribute only 95 percent of the total dose. 'Ihe contribution frun the other isotopes in the mix are accounted for by then nultiplying the dose calculated in equation 49 by a factor of 1.05.

'Ihe total dose is then normalized against the Appendix I guidance:

Total Dose = 1.05

• dose (mrem)/ 3.0 (mrem)

(54)

In IEBOG the Ai.4, j vnlua3 were taken frcn ODCM inplementing precedure AP.310-2L, Revision 2.

'Ihose numbers are derived from the KCs IADIAP liquid effluent dose code (modified with site specific data) which does give a conservative estimate of the dose.

PISUL'IS Appendix B :ontains a table showing the results of the IEBOG code as describcd above.

For the plant conditions at shutdown, namely a 0.05 percent failed fuel and a 0.04 to 0.07 gpn primary to secordary leak rate, IEIOG indicates that the offsite dose frca liquid effluents would be roughly 40 percent of Appendix I guidan At that level of fuel failure, a high probability of meeting Appendix I exists for leakages on the order of 0.2 gpn.

Higher levels of failure would reduce that margin.

For exarple, at 0.08 percent failed fuel, a continuous leakage of 0.10 to 0.15 gpu may challenge us to meet Appendix I.

12 The line on the table divides the h below A@endix I frc those above.

The box indicates the case that was evaluated in the verification.

CONCIUSIONS While the operational envelope for primary to secxxrlary leakage ard failed fuel fraction predicted by LEBOG is considerably narrower than the limits given in the Technical Specifications, the "boundary" developed in this evaluation should not be wastrued as an instantaneous cperational limit since IEBOG expressly assumes average, steady state plant perfonnan The IEBOG results show throtqh conservative analysis that the plant could operate arri meet Appendix I with the current plant configuration. Operational flexibility exists to make other minor adjustments to the plant equipnent or its operating procedures to mitigate offsite dose if the Appendix I objectives are appr : ached.

In addition to performinJ a preliminary evaluaticn of the liquid effluent related systems, IEEG can be adapted to predict (online) the Appendix I inplicaticos of various plant operations and assist in the developnent of corrective actions.

1

e D

APETNDIX A:

Befeltmoes

(3) Oconee Radioctweintry dW Pwgcus, cemiannual Report (January-June 1974);

Appendix A:

Fission Product Spiking by P.J. Grant, with Rahk and Wilcox (2) Rancho Seco Nuclear Generating Station Updated Safety Analysis Report (USAR)

(3) Rancho Seco Offsite Dose Calculation Manual (ODCM), Revision 2 (4) The Radiological Health Handbook, ccurpiled by Shleien and Terpilaka (Nucleaon Isctern Associates,1984)

(5)

Industrywide Nuclear Power Plant Performance Indicators:

1987 Mid-year Report, published by the Institute of Nuclear Power Operations (6) Chart of the Nuclides, prepared by the Knolls Atcznic Power Iaboratory (1977)

(7) Rancho Seco Nuclear Generating Station: Systems Training Manuals, prepared by the General Physics Cbrporation (1983)

(8) Calculation of Relaaaaa of Radioactiva Materials in naaarus and Liquid Effluents frtan Pressurized Water Reactors (IMR-GAIE code) NURD3/CR-0017 rev.1, by Chandrasekaran, Ise, and Willis, 1985 (9)

In-Plant Scurce 7brm Meast.uumnts at Rancho Seco Station, NURIr-/CR-2348, by III and G Idaho, Inc. and Exxon Nuclear Idaho Co, Inc. 1981 (10) Rancho Seco Tedinical Specification Manual (Docket 50-312)

(11)

Introduction to Nuclear EngineerirxJ by IaMarsh (12)

Bases for Icwer Limit of Detection Values for Rancho Seco Liquid Effluents, prgared by R. Oesterling (United Energy Services), June 1987 (13)

Proposal for Isase of Dual Sluiceable Demineralizer system and Technical Services (RFP 86-08) prepartd by Pacific 1Arlear Syst=sts Inc., % =har 1986.

(14)

RaWk and Wilocx Draft Ebel Failure Report, prepared for t!e Oore Perforrance Ctr:nittee of the B&W Owners Group, by DL Husser, 22 Sept, 1986 A-1 l

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e (15) NRC Inspection Report 50-312/86-15 ocnheted April-May 1986, by Ha e ard vuhas (16) Reactor Coolant Activit*/ Trendity Graphs, prepared by the Nuclear Chanist for the years 1984-1985 I

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4 APPENDIX B:

Code Results l

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RANCHO SEC0 OPERATIONAL' ENVELOPE -- FRACTION OF APPENDIX I ALARA GUIDANCE FOR GIVEN FUEL DEFECT AND PRIMARY TO SECONDARY LEAKAGE AVERAGE PRI-SEC LEAK RATE (GPhi FAILED FUEL (%) 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50

.01 0.28.

0.32 0.36 0.40 0.44 0.48 0.52 0.56 0.60 0.64

.02 0.32 0.40 0.48 0.56 0.64 0.72 0.79 0.87 0.95 1.03

.03 0.36 0.48 0.60 0.72 0.83 0.95 1.07 1.19 1.30 1.42

.04 0.40 0.56 0.72 0.87 1.03 1.19 1.34 1.50 1.66 1.81

.05 0.44 0.64 0.83 1.03 1.22 1,42 1.62 1.81 2.01 2.20

.06 0.48 0.72 0.95 1.19 1.42 1.66 1.89 2.13 2.36 2.60

.07 0.52 0.79 1.07 1.34 1.62 1.89 2.17 2.44 2.71 2.99

.08 0.56 0.87 1.19 1.50 1.81 2.13 2.44 2.75 3.07 3.38

.09 0.60 0.95 1.30 1.66 2.01 2.36 2.71 3.07 3.42 3.77

.10 0.64 1.03 1.42 1.81 2.20 2.60 2.99 3.38 3.77 4.16

.11 0.68 1.11 1.54 1.97 2.40 2.83 3.26 3.69 4.12 4.56

.12 0.72 1.19 1.66 2.13 2.60 3.07 3.54 4.01 4.48 4.95

.13 0.75 1.26 1.77 2.28 2.79 3.30 3.81 4.32 4.83 5.34

.14 0.79 1.34 1.89 2.44 2.99 3.54 4.09 4.63 5.18 5.73

.15 0.83 1.42 2.01 2.60 3.18 3.77 4.36 4.95 5.54 6.12 ASSUMPTIONS:

l FUEL ESCAPE COEFFICIENT 1E-08 1/SEC RCS & SECONDARY VOLUMES 88200 / 216000 GALS RESPECTIVELY LETDOWN FLOW RATE 45 GPM PURIFICATION DEMIN OF 100 CONDENSATE POLISHER DEMIN OF 10 FRACTION PROCESSED IN SKIDS.5 SLUICE DEMIN SKIDS DF 50 PLANT AVAILABILITY.61 t

PRIMARY CS-137/I-131 RATIO.14 l

CS-134/CS-137 RATIO IN RHUTS.55 l

DISCHARGE FLOW (W/ DILUTION) 8496.409 GPM l

ESTIMATED TRITIUM 100 CI/ YEAR TOTAL SECONDARY SIDE LOSSES 50 GPM SECONDARY LEAKAGE DIRECTED TO RHUT 50 GPM l

VOLUME TRANSFERRED FROM DFCST 1000000 GAL /YR B-1 1

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y APPENDIX C:

Text of I.EBOG l

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i 10 REM: PROGRAM LEBOL 20 REM: ASSISTS IN THE DEVELOPMENT OF PLANT OPERATING 30 REM: LIMITS AT RANCHO SECO BASED ON OFFSITE DOSES 40 REM: RESULTING FROM RADI0 ACTIVE LIQUID EFFLVENT 70 DEFINT X, Z 80 DIM 00SE(20,20) 90 REM:

100 REM: PLANT /IS0 TOPE DATA 110 READ ESCAPE, FISS, POWER, YlELD 120 DATA 1.0E-8, 3.3E16, 2772,.029 130 T0P-ESCAPE *FISS* POWER

• YIELD 140 READ VOLRCS, LETDOWN, DECAY, DFl 180 DATA 8.82E4,.75, 1.0E-6, 100 185 CLEANVP= LETDOWN *(1-(1/DF1))/V0LRCS 190 BOTTOM-V0LRCS*3785*37000l*(CLEANVP+ DECAY) 200 READ CIRATIO, CCRATIO, AVAIL, DAYS 205 DATA.140,.55,.61, 365 210 TIME-DAYS *24*60* AVAIL 220 READ VOLSEC, CONDFLO, DF2, DFLOW, DRCST 225 DATA 2.16E5, 1.65E4, 10., 18.93, IE6 230 DEMIN-CONDFLO*(1-(1/DF2))/VOLSEC 235 READ AIJ137, AIJ134, AIJH3, APPI, H3 240 DATA 5.57E3, 8.93E3, 9.27E-2, 3.0, 100, 245 READ LOSSES, SECLEAK, FRACTION, DF3 250 DATA 50, 50,.50, 50.

280 READ FSTART, FEND, FSTEP, LSTART, LEND, LSTEP 285 DATA.0001,.0015,.0001,.05,.50,.05 290 XEND 1+(FEND-FSTART)/FSTEP 295 ZEND-1+(LEND LSTART)/LSTEP 298 LENDIT-LEND + LSTEP 300 REM:

310 REM: BEGIN FAILED FUEL FRACTION LOOP 320 FOR FFUEL-FSTART TO FEND STEP FSTEP 330 X-1 +(FFUEL-FSTART)/FSTEP 340 1131-FFUEL*(T0P/ BOTTOM)*.000001 350 PRICS137-CIRAT!0*Il31 400 REM:

410 REM: BEGIN PRI-SEC LEAK LOOP 420 FOR LEAK-LSTART TO LENDIT STEP LSTEP 430 Z-1+(LEAK-LSTART)/LSTEP 440 CSRATE-LEAK *3785*PRICS137* AVAIL 445 CONC- (CSRATE/VOLSEC)/(DEMIN+(LOSSES /VOLSEC))

450 SECCS137-CONC *SECLEAK* TIME 460 REM:

470 REM: SLUICE DEMIN/RADWASTE OPERATION 500 SLVICE-SECCS137* FRACTION 510 NOSLVICE-SECCS137*(1-FRACTION) 550 EFFCS137-N0 SLUICE + SLVICE/DF3 560 CS134-EFFCS137*CCRATIO 570 D1-EFFCS137*AIJ137 575 02-CS134*A!J134 580 D3 H3*AIJH3 590 04-DRCST*.000000l*3785*.00000l*10500!

600 DOSE (X,Z)- (Dl+02+03+D4)/DFLOW 610 NEXT LEAK 620 NEXT FFUEL 630 REM:

640 REM:

650 REM:

660 REM:

C-1 t

670 REM:

675 REM:

680 REM:

685 REM:

690 REM:

695 REM:

700 REM: OUTPUT 710 LPRINT "RANCHO SECO OPERATIONAL ENVELOPE -- FRACTION OF APPENDIX I ALARA" 720 LPRINT "GUIDANCE FOR GIVEN FUEL DEFECT AND PRIMARY TO SECONDARY LEAKAGE" 740 LPRINT CHR$(10) 760 LPRINT TAB (19), "AVERAGE PRI-SEC LEAK RATE (GPM)"

780 LPRINT "FAILE0" 785 LPRINT "FUEL (%)";

790 FOR LEAK-LSTART TO LEND STEP LSTEP

-800 LPRINT USING " #.## '; LEAK;

NEXT LEAK 900 LPRINT CHR$(10) 910 LPRINT "----------------------------- ---------------------------- "

920 FOR X-1 TO XEND 930 LPRINT CHR$(10) 935 LPRINT USING ".##

"; X*FSTEP*1001; 940 FOR Z TO ZEND 950 VALUE-1.05*00SE(X,Z)/APPI 960 LPRINT USING "

1. 1. .##"; VALUE;

NEXT Z 970 NEXT X 975 LPRINT "

LPRINT "

980 LPRINT "------------------------------------------------- -------- "

1000 REM: OUTPUT ASSUMPTIONS 1010 LPRINT "

1020 LPRINT "ASSUMPTIONS:"

1025 LPRINT "FUEL ESCAPE COEFFICIENT"; ESCAPE; " 1/SEC" 1030 LPRINT "RCS & SECONDARY VOLUMES";VOLRCS;"/";VOLSEC;"GALS RESPECTIVELY" 1040 LPRINT "LETDOWN FLOW RATE"; LETDOWN *60; "GPM" 1050 LPRINT "PURIFivATION DEMIN OF"; 0F1 1060 LPRINT "CONDENSATE POLISHER DEMIN OF"; 0F2 1070 LPRINT "FRACTION PROCESSED IN SKIDS"; FRACTION 1080 LPRINT "SLUICE DEMIN SKIDS OF"; DF3 1090 LPRINT "PLANT AVAILABILITY"; AVAIL 1100 LPRINT "PRIMARY CS-137/I-131 RATIO"; CIRATIO 1110 LPRINT "CS-134/CS-137 RATIO IN RHUTS";CCRATIO 1120 LPRINT "DISCHARGE FLOW (W/ DILUTION)"; 0 FLOW *448.833;" GPM" 1130 LPRINT "ESTIMATED TRITIUM"; H3; " CI/ YEAR" 1140 LPRINT "TOTAL SECONDARY SIDE LOSSES"; LOSSES; " GPM" 1150 LPRINT "SECONDARY LEAKAGE DIRECTED TO RHUT"; SECLEAK;" GPM" 1160 LPRINT "VOLUME TRANSFERRED FROM DRCST";0RCST;" GAL /YR" 9999 END i

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APPD OIX D:

Ncmenclatum

o me following list contains a definiticn for each of the variables used in the LEB0G ccuputer code.

Se variables are presented in the order in which they appear in the text.

ESCAPE = 'Ihe escape coefficient for I-131 da<v ribing the rate at which iodine leaves the fuel and enters the fuel red gap.

In the prsmiire discussion it was referred to as vi. S e value is 1E-8 sec-1 FISS=

A constant relating the rumber of fissions to the thermal energy produced. h value of 3.3E16 fissions /W-sec ccnes frca Reference 1.

POWER = h rated thermal pcwer of the reactor.

It is defined as 2772 MR in the Rancho Seco 'Ibchnical Specificacions (Reference 10).

YIELD = S e fission yield of I-131 expressed as a fraction. S e value of.029 was fourd in Reference 1.

'IOPm A durtfly variable (intermediate calculational Etep).

CI N 2e clean-up coefficient of the primary system for I-131 calculated in units of sec-1 BCfrICH= A dumy variabla (inta-diate calculational step)

VOIRCS= Se volume of water in the primary coolant. USAR AEpendix 14D gives a value of 3.34E8 ml which equates roughly to 88,200 gallcos.

IEI1XW= h letdown flowrate thrcugh the Puriflcstion and Intdown (FIS) systca was set at 45 gpn per USAR and confirmed by plant cperations staff.

DECAY = me decay constant for I-131.

Se value of 1E-6 sec~1 ccnes frtn the chart of the nuclides (Reference 6).

DFl= Ae damntamination factor for the PIS for I-131 was set at a 100 per in-plant measurement (Referen 9).

21s assumes operation of the purification E neralizer; no credit is taken for the coolant radwaste i

primary or secondary demin processing.

l CIRATIO= Se ratio of Cs-137 over I-131 in the primary coolant determined to be.14 based cn historical data (average 1984).

CCSATIO= Se ratio of Cs-134 over Cs-137 in the Regenerative ibld42p tanks, h value of.55 (average) ccnes frtn the Radiation Prota: tion data base manager ccupilation of all RHUT releases.

AVAIIs 'Ihe plant availability was set conservatively at the industry average for PWRs over the last seven years as p2blished by DTIO.

DAYS = Se tima period of interest (365 days).

D-1 1

i TIME = 'Ibe time period of intemst in minutes adjusted for plant availability.

VOISEO=

'Ihe volume. of the secordary system.

'Ibe value of 2.16E5 gal omes frm an in-house best estimate by cperations staff.

CONDFII>

'Ihe average flow rate of camndary water through the cxmdensate polishers.

'Ibe value of 1.65E4 gpn cmes fra the General Physics training manuals (Reference 7, Chapter 20).

DF2= 'Ihe decontamination factor for the condensate polishers.

'Ihe value of 100 c aes frun the GAIE code assunptions (Reference 8) for Radwaste demineralizer (H+ OH-type); this was cut by a factor of 10 (to a value of 10) for conservatism.

DFLOW=

'Ihe average dilution flow provided by the total plant effluent as measured by FE 95108.

By administrative procedure the mininum flcw is set at 11.1 cubic feet per secorri (5000 gpm).

Plant management intends to increase that minimum to 18.93 cfs (8500 gpn) in 1988, and this value was used.

[RGir= 'Ibe annual amount of primary side water (in gals) expected to be transferred to the RHUIS for release offsite in liquid effluents.

DEMDb 1he removal rate coefficient for the ocxidensate polishers calculated in units of min-l.

AIJ137= 'Ihe liquid effluent dose @ ion factor for Cs-137.

'Ihe value of 3

5.57E3 mrem (adult whole body)-ft /Ci-sec cmes frm the ODCM (Reference 3)

AIJ134= 'Ihe liquid effluent d.ce-dispersion factor for Cs-134.

'Ihe value of 3

3.93E3 mrem (adult whole body)- ft / Ci-sec caes frm the ODCM (referen 3).

AIJH3= 'Ibe liquid effluent dose-dispersicr, factor for H-3.

'Ihe value of 3

9.27E-2 mrem (adult whole body)- ft / Ci-sec ccnes frm the ODCM (reference 3).

APPI= 'Ihe offsite dose objective given in 10 CFR 50 Appendix I (3 mraVyr).

H3= 'Ihe estimated amount of tritium released anrually in liquid effluents in Curies; the value of 100 is di m m W in section III.d.

l If6SES= 'Ibe average of the total leakage rate in gpn frm the secondary sida.

1 For conservatism, it is taken as the same ancunt of leakage reaching the RHUr (ie SEC2EAK), that is, 50 gpn.

SEC1EAK=

'Ihe average rate of secondary leakage which reaches the RHUr for eventual release offsite. As di m mcad in section III.d, the value is 50 gpn.

FRACTICtb

'Ibe fraction of RHUr water pr-cad through the sluiceable demineralizer skids. Since the rate of effluent generation has been estimated D-2 l

at 86 gpn and the systen has a ncaninal capacity of 100 gpn, a value of 1 could be used. However, a value of.5 was used for conservatism.

DF3=

'1he hotamination factor for the sluiceable hineralizer skids.

A value of 1000 was given in the Pacific Nuclear bid proposal (Reference 13),

and 50 was used for conservatism.

XDh> h number of failed fuel increments.

ZDO= 'Ibe number of leak rate iruwwits.

IINDIT= A counter for the number of leak rate irmasits.

FS'IARP, FEND, FSTEP= 'Ibe beginning, end, and increment of the range of failed fuel fraction considered in the ocmpiter run.

ISIART, IIND, IETEP=

'Ihe beginning, end, and irma. west of the range of primary to secondary leak rate.

X=

A counter for the number of failed fuel incrunents.

Z= A counter for the number of leak rate ina w &its.

I131= 'Ihe concentration of I-131 (uCi/ml) in the primary systan calculated for a given failed fuel fracticn.

PRICS137=

'Ihe concentraticn of Cs-137 in the primary system calculated in units of Curies per milliliter.

CSRATD= 'Ibe rate of transfer of Cs-137 activity to the namndary systen through the OISG leak in units of Curies per minute.

CONO= 'Ihe equilibrium con ntraticn of Cs-137 in the secondary side in units of C1/ gal.

SECCS137= 'Ihe amount of Cs-137 that reaches the receiving IHJr, excludirg prirmy side transfers, in curies.

SI1JICE= Of the fraction of RHUr water pr-W through the sluice hins, the amount of CS-137 in Curies left in that fraction after derpin operaticn.

NOSIIJICD= 'Ibe amount of CS-137 in the RHUr water not pmW through the sluice h ins in Curies.

11 i " - 37= 'Ihe amount of Cs-137 to be released offsite in Curies,

.ne anmnt of Cs-134 to be relcased offsite (Curies) raticed frun the i '

u C2 1.

curoo term.

D-3

Di=

'Ibe Cs-137 source term fran secondary leakage ultiplied by its Aij 3

value, units of mrem *ft /sec.

D2=

'lhe Cs-134 sourm tenu fran secondary leakage ultiplied by its Aij 3

value, units of mran*ft /sec.

The total H-3 source term festensibly frun IRCSI transfers) multiplied by D3=

its Aij value, units of mrem *ft /sec.

Di=

The Cs-137 acurce term frun primary side transfers nultiplied by an mren*ft /sec. j value to include the prtportionate amount of Cs-134, units of effective Ai 3

DOSE (X,Z)= "he adult whole body dose (in mrum) resulting frun the release of the sources hibed in D1, D2, D3, and D4 into the liquid effluent for a given failed fuel fraction represented by X and primary to secondary leakage represented by Z.

VAIDE= The offsite dose normalized against the Appendix I number and biaW upward by 5 percent to account for the effect of other isotopes. The fraction has no units.

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Verification i

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'Ibe IEBOG code was verified by hand calculation for the case of a 0.05 percent failed fuel and a continuous 0.10 gpn primary to secondary leak rate.

'Ihe process was h==nced in Calculation EP-lEBOG, its subparts, and supporting calculations. 'Ibey can be sumarized as follows:

EP-IEBOG-!RIN:

'Ibe inplementation of the methods detailed in the basis h==nt.

EP-IEBOG-SUB1:

Determination of the ratio of Cs-137 to I-131 in the primary coolant using USAR data.

EP-IEB3G-SUB2:

Demonstrates that the historical ratio (1981-1986) of Cs-134 to Cs-137 in the liquid effluent has been 55 percent.

D=ainimi.x-- tes that the adult whole body dose resulting frca EP-IEBOG-SUB3 :

a oesium in the effluent is more restrictive than the thyroid organ dose produced by the release of I-131.

EP-LEBOG-SUB4:

Calculates the average plant availability for IHRs in the Unites States during the years 1980-1986 hamai on INN data.

EP-IEBOG-SUB5:

Calculates the fuel escape coefficient to make the IEB3G prediction of I-131 concentration in the primary coolant consistent with the USAR analysis.

EP-IEB3G-SUB6:

Investigates the consequences of seeing less than 100 as the damntamination factor for the Sluiceable Demineralizers and haviry to run that equignent in the less efficient recirculation configuration.

EP-IIB 3G-SUB7 : Determines the average ratio of Cs-137 to I-131 in the primary coolant near equilibrium at full power using historical data.

EP-IID-SUB4 :

Calculates the average annual release of tritium in liquid effluent.

EP-Q H7T-SUB2:

Frun operations logs, determines the annual volume of demineralized water pr-cal through the DM systan.

'Ihis documentaticn is on file in the Enviidme.ntal ', Mon Depart: ment, RS-20.

E-1

.A APPENDIX F Sumary of Revisions I

'1his Appeniix sumarizes the charges mde to the IIBCG (Liquid Effluent h aad Operating Guides) ccatputer code developed at Ranch Seco to estimate the ranges of failed fuel and primary to secondary leakage over which ccmpliance with Appendix I can be maintained.

'Ihe first draft was issued for ccanent in November of 1987.

REVISION 1 ( M r, 1987):

(1)

'Ihe draft code uses the value of 2E-8 soc-1 r - nded in the USAR for the escape coefficient for :-131 entering the gap frun the fuel un$er accident conditions *

'Ibe value of 1E-8 sec-1 used in the Revision 2 accounts for the differences in methodology between the USAR and IEB3G; it was derived to insure that the concentration of I-131 predicted in IEB3G was censistant with that provided in the USAR.

(2)

'Ibe Reactor Coolant System volume in the draft (68,000 gals) was calculated based on an assunpt. ion in the USAR regarding the relative rate of letdown to the Purification and Makeup Systern.

'Ihe actual value fourd elsewhere in the USAR is 3.34E8 ml or rotqhly 88,200 gals.

'Ihis value is consistent with estimation of the site chemistry staff.

(3)

In spite of the Pacific Nuclear prcposal to bid which estimated 1000 as the DF across the sluiceable demineralizer skids, many meubers of our staff felt that this number was overly optimistic.

Consequently, the final uses a value of 100 for conservatism.

Also, scana consideration is given to the possibility of a DF of only 10 for this equipnent.

(4) Recent staff meetirgs have raised concerns about running the sluice demin skids in the "straight-through" made described in IEBOG because of the lack of control over the inlet water chemistry.

An additional subcalculation was included to examine the use of having the sluice demins in the less efficient "recirculation" rede.

(5)

}ere parametric study was done regarding the effect of the three major demineralizers.

(6) Generally, rcre di== ion was added to better illuminate the assunptions used ard the different sluice demin skids configurations and its range of efficiency.

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REVISION 2 (January,1988)

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(1)

Previcus versions used a ratio of Cs-137 to I-131 in the primary coolant based on USAR data for expected failed fuel conditions. Based on NRC concerns i

about emmive conservatism in the USAR values, a new ratio was calculated based on historical data at Rancho Seco.

'Ihe ratio was reduced by approximately a factor of 50.

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(2)

Instead of estimating the amount of activity in the RHL7Fs by perfonning an activity balance on the seandary mnsidering only primary inleakage and removal by the condensate Polishers, a new differential equation was derived to hibe the concentraticn of the secondary at equilibrium.

Activity directed to the RHLTrs was estimated by nultiplying this concentration by the anticipated volume of secondary side leakage.

(3)

'Ibe decontamination factors for the Polishers and the Sluice Domins ware reduced fr m 100 and 100 to 10 and 50, respectively.

(4)

'Ihis revision includes the doses resulting frm the transfer of tritium and cesium contaminated water frm the primary side (Demineralized Reactor Cbolant Storage Tank or CRCST).

(5) 'Ihe parametric study was deleted.

(6) 'Ibe period of interest was changed frm a quarter to a full year.

(7) 'Ihe average dilution flow offsite was increased frm 5000 to 8500 gpm.

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