Text

Chapter 11, Radwaste Mgt, to CESSAR Sys 80+ Std Design
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7 (Sheet 1 of 2)

CESSAR !!n%uict O

EFFECTIVE PAGE LISTING CHAPTER 11 Table of contents page Amendment i

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Pace Amendment 11.1-1 E

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11.1-13 E-11.1-14 E

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11.1-17 E

11.1-18 E

11.1-19 E

11.1-20 11.2-1 E

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11.3-1 E

11.3-2 E

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EFJECTIVE PAGE LISTING (Cont'd)

CHAPTER 11 Text (Cont'd)

P_a g e Amendment 11.3-4 E

11.3-5 E

11.4-1 E

11.4-2 E

11.4-3 E

11.4-4 E

11.5-1 E

Tables Amendment 11.1.1-1 (Sheet 1)

E 11.1.1-1 (Sheet 2)

E 11.1.1-2 E

11.1.1-3 E

j 11.1.2-1 11.1.2-2 (Sheet 1) 11.1.2-2 (Sheet 2) 11.1.2-3 11.1.2-4 11.1.2-5 (Sheet 1) 11.1.2-5 (Sheet 2) 11.1.2-6 E

11.1.2-7 E

11.1.2-8 E

11.1.2-9 11.1.3-1 11.1.3-2 E

11.1.3-3 E

11.1.3-4 (Sheet 1) 11.1.3-4 (Sheet 2) 11.1.3-4 (Sheet 3) 11.1.6-1 11.1.8-1 E

11.1.9-1 E

11.1.9-2 E

11.3.4-1 E

O Amendment E December 30, 1988

CESSAR E!%ncam.

O k

TABLE OF CONTENTS CHAPTER 11 Section Subiect Pace No.

11.0 RADIOACTIVE WASTE MANAGEMENT 11.1-1 11.1 SOURCE 11.1-1 l

11.1.1 DESIGN BASIS SOURCE TERMS 11.1-1 11.1.1.1 Maximum Fission Product Activities 11.1-1 3

in Reactor Coolant 11.1.1.2 Normal Operatina Source Terms 11.1-3 Includina Anticipated Operational Occurrences

]

11.1.2 DEPOSITED CRUD ACTIVITIES 11.1-5 11.1.3 TRITIUM PRODUCTION IN REACTOR COOLANT 11.1-9 11.1.3.1 Activation Sources of Tritium 11.1-9

\\m 11.1.3.2 Tritium From Fission 11.1-10 11.1.4 NEUTRON ACTIVATION PRODUCTS 11.1-11 11.1.4.1 Nitrocen-16 Activity 11.1-11 11.1.4.2 Carbon-14 Production 11.1-11 11.1.5 FUEL EXPERIENCE 11.1-12 11.1.6 LEAKAGE SOURCES 11.1-12 11.1.7 SPENT FUEL POOL FISSION PRODUCT AND 11.1-13 CORR 0SION PRODUCT ACTIVITIES 11.1.8 STEAM GENERATOR ACTIVITY MODEL 11.1-13 11.1.9 RADIOACTIVE WASTE SYSTEMS 11.1-15 11.2 LIOUID WASTE MANAGEMENT SYSTEM 11.2-1 11.2.1 DESIGN BASES 11.2-1 E

11.2.1.1 Criteria 11.2-1 Amendment E i

December 30, 1988

l k

ER IC ATIT,N i

O TABLE OF CONTENTS (Cont'd)

CHAPTER 11 Section Subiect Pace No.

11.2.1.2 Codes and Standards 11.2-1 11.2.1 7 Features 11.2-1 11.2.2 SYSTEM DESCRIPTION 11.2-2 11.2.2.1 General Description 11.2-2 11.2.2.2 Components Descriptions 11.2-3 11.2.2.2.1 Waste Collection Tanks 11.2-3 11.2.2.2.2 Waste Sample Tanks 11.2-3 11.2.2.2.3 Process Pumps 11.2-3 E

11.2.2.2.4 Process Vessels 11.2-3 11.2.2.2.5 Provisions for Mobile Equipment 11.2-4 11.2.2.2.6 Steam Generator Drain Tank 11.2-4 11.2.2.3 System Oparation 11.2-4 11.2.3 SAFETY EVALUATION 11.2-5 11.2.4 INSPECTION AND TESTING REQUIREMENTS 11.2-5 11.2.5 INSTRUMENTATION REQUIREMENTS 11.2-6 11.3 GASEOUS WASTE MANAGEMENT SYSTEM 11.3-1 11.3.1 DESIGN BASES 11.3-1 I

11.3.1.1 Criteria 11.3-1 11.3.1.2 Codes and Standards 11.3-2 l

11.3.2 SYSTEM DESCRIPTION 11.3-2 11.3.2.1 General Description 11.3-2 11.3.2.2 Components Description 11.3-4 11.3.2.2.1 Charcoal Beds 11.3-4 11.3.2.2.2 Gas Dryers 11.3-4 11.3.2.2.3 Piping and Valves 11.3-4 11.3.3 INSPECTION AND TESTING REQUIREMENTS 11.3-5 Amendment E ii December 30, 1988

CESSAR nnincarcu l'%

t i

G' TABLE OF CONTENTS (Cont'd)

CHAPTER 11 Section Subiect Pace No.

11.3.4 INSTRUMENTATION REQUIREMENTS 11.3-5 11.4 SOLID WASTE MANAGEMENT SYSTEM 11.4-1 11.4.1 DESIGN BASES 11.4-1 11.4.1.1 Criteria

.'.1.4-1 11.4.1.2 Codes and Standards 11.4-1 11.4.1.3 Features 11.4-1 11.4.2 SYSTEM DESCRIPTION 11.4-2 11.4.2.1 General Description 11.4-2 11.4.2.2 Comoonents Description 11.4-2 11.4.2.2.1 HIC F.11/ Dewatering Head 11.4-2 11.4.2.2.2 Slurry Pump 11.4-3 11.4.2.2.3 Radwaste Building Crane 11.4-3 11.4.2.2.4 Dry Solids Compactor 11.4-3 11.4.2.3 System Operation 11.:-3 11,4.3 SAFETY EVALUATION 11.?-4 11.4.4 INSPECTION AND TESTING REQUIREMENTS 11.4-4 11.4.5 INSTRUMENTATION. REQUIREMENTS 11.4-4 11.5 PROCESS AND EFFLUENT RADIOLOGICAL 11.5-1 MONITORING AND SAMPLING SYSTEMS APPENDIX 11A CORE RESIDENCE TIMES 11A-1 Ob Amendment E iii December 30, 1988

CESSAR EnFlbuou O

LIST OF TABLES CHAPTER 11 Table Sub'i ect 11.1.1-1 Basis for Reactor Coolant Fission Product Activities 11.1.1-2 Maximum Activities in the Reactor Coolant Due to Continuous Operation at Maximum Power with One E

Percent Failed Fuel 11.1.1-3 Reactor Coolant System Activities During Normal Operations Including Anticipated Operational occurrences 11.1.2-1 Long-Lived Isotopes in Crud 11.1.

2 Measured Radioactive Crud Activity (dpm/mg-crud) 11.1.2 3 System Parameters 11.1.2 1 System Parameters 11.1.2-5 Average and Minimum Residence Times, Days 11.1.2-6 Assumed System Parameters, System 80 11.1.2-7 Long-Lived Crud Activity for a Standard 3817 Mwt Plant 11.1.2-8 Average Calculated Reactor Coolant Crud Activity 11.1.2-9 Equilibrium Crud Film Thickness l

l 11.1.3-1 Tritium Activation Reactions 11.1.3-2 Parameters Which Influence Tritium Production Determination 11.1.3-3 Conservative Estimate for Tritium Production in Reactor Coolant 11.1.3-4 Tritium Production and Release at Operating PWRs 11.1.6-1 Leakage Assumptions From C-E Supplied Equipment E

1 Amendment E iv December 30, 1988

n-CESSAR !anncan, i

\\

LIST OF TABLES (Cont'd)

CHAPTER 11 Table Subject 11.1.8-1 Basis for Steam Generator Liquid Activities 31.1.9-1 Annual Spent Resin Activity Input to SWMS

.11.1.9-2 Specific Activities of Sources to the GWMS During Normal Operation 11.3.4-1 GWMS Instrumentation and Control E

1 Amendment E v

December 30, 1988

m CESSAR En&"icarien nU 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE 11.1.1 DESIGN BASIS' SOURCE TERMS 11.1.1.1 Maximum Fission Product Activities in Reactor Coolant l

Maximum fission product activities will be used as design basis I

source terms for shielding and facilities design.

Source terms for calculating the consequences of postulated accidents are discussed in Chapter 15.

The isotopes chosen for consideration E.

in the maximum case are those which are significant for design purposes by reason of a combination of energy, half-life or abundance.

The mathematical model used to determine the ' concentration of nuclides in the Reactor Coolant System involves a group of linear, first order differential equations.

These equations are obtained by applying a mass balance for production and removal for the fuel pellet region as well'as the coolant region.

In the

(

fuel pellet region, the mass balance includes fission product I

production by direct fission yield, by parent fise. ion product s

decay and by neutron activation while the removal includes-decay, E

neutron activation and escape from the pellet.

In the coolant region the analysis includes the fission product production by l

cscape from the fuel through defective fuel rod cladding, parent decay in the coolant and neutron activation of coolant fission products.

Removal is by decay, by coolant purification, by feed and bleed operations (for fuel burnup), by leakage and other feed and biced operations such as startups and shutdowns as well as load follow operation.

The expression derived to determine the fission product inventory in the fuel pellet region is:

dN

= (F)(Y )(P) + (f,3 A,3)Np,4,3 + a3 4N g

j j

p,3

-(Aj+Vi + oj d)Np,j (1)

The expression derived to determine the fission product inventory in the reactor coolant region is:

dN

" (D)(Vj)(N j) + (f,3 A,3)Nc,i-1 + ("j d.

)N j

j c,j

.\\

Amendment E l

11.1-1 December 30, 1988 I

ke!b!h!h k!I bb T IC ATl!N O

(1 nj)C

+h)N

-(Aj+

nj + C c,i

- Ct (2) 0 where the variables are identified as:

N

= Population, atoms F

= Average fission rate, fissions /Mwt - sec Y

= U-235 fission yield of nuclide, fraction (Reference 1)

P

= Core power, Mwt

-1 A

= Decay constant, src (Reference 2)

J

= Microscopic capture cross section cm (Reference 3) o p

4

= Thermal neutron flux, neutrong/cm -sec v

= Escape rate coefficient, sec f

= Branching fraction t

= Time, seconds D

= Defective fuel cladding, fraction CVR

= Core coolant volume to reactor coolant volume ratio, fraction b

= CVCS purification flow rate during power operation, lbm/sec

'4

= Reactor Coolant System mass during power operation, lbm

= Rosin efficiency of CVCS lon exchanger and gas stripper efficiency (subscripted for a particular nuclide)

C

= Beginning of core life boron concentration, ppm g

C

= Boron concentration reduction rate because of feed and bleed, ppm /sec L

= Leakage or other feed and bleed from the reactor coolant, lbm/sec and where the subscripts are identified as:

th i

=i nuclide

=precursortoifh nuclide for decay i-1 j

= precursor to i nuclide for neutron activation p

= pellet region c

= coolant region E

Escape rate coefficients are used to represent the overall release from the fuel pellets to the coolant.

The escape rate coefficient is an empirical value which was derived from experiments initiated by Bettis and run in the NRX and MTR reactors (Reference 4).

The escape rate coefficients were Amendment E 11.1-2 December 30, 1988

CESSAR EMnncm.

l DU obtained from test rods which were operated at high linear heat rates.

The linear heat rates were uniform over the test sections of 10.25 inches in length. 'The exact linear heat rates were not precisely known but post-irradiation inspection showed that some l

test specimens had experienced centerline melting.

Later tests were done in Canada to determine the effect of rod length on the release of fission gases and iodines from defective fuel rods (Reference 5).

A by-product of these experiments was the 1

relationship between linear heat rate and escape rate coefficient.

The average heat rate for a fuel rod is well belowlE the values that correspond to the selected escape rate coefficients for halogens and noble gases.

The presently used l E escape rate coefficients are conservative.

These escape. rate coefficients are based on a linear heat rate coefficient of 18 kw/ft rather than the design maximum linear heat rate of 21 kw/ft.

Shown in Table 11.1.1-1 are the values of bounding parameters E

used to evaluate the maximum reactor coolant fission product activities.

The maximum activities are presented in Table 11.1.1-2.

O 11.1.1.2 Normal Operatinc Source Terms Includina Anticipated Operational Occurrences The data in Table 11.1.1-3 represent the expected normal fission product activities for system operational performance calculations.

The activities for this case are evaluated based E

I on the bounding performance parameters from Table 11.1.1-1.

i nU Amendment E 11.1-3 December 30, 1988

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l f

CESSAR !!!nneuion O\\

THIS PAGE INTENTIONALLY BLANK O

O 11.1-4

CESS.AR EMUICATl!N oV TABLE 11.1.1-1 (Sheet I of 2)

BASIS FOR REACTOR COOLANT FISSION PRODUCT ACTIVITIES PARAMETER MAXIMUM PERFORMANCE (I)'

)

Core Power Level (Mwt) 4100 3800 Duration of Reactor Operation (core cycles) 5 5

Equilibrium Fuel Cycle (Equivalent Fuel 292 292 Power Days)

E Average Thermal Fission Rate 3.10x10 3.lcX10

]

16 16 (Fission /MW-second) 13 13 ThermglNeutronFlux-average 5.50x10 5.5X10 (n/cm -second)

Fraction of Failed Fuel 0.0025 0.0025 5

5 Reactor Coolant Mass including pressurizer 5.752x10 5.713x10 j

(Pounds)

E Core Coolant Volume to Reactor Coolant 0.0723 0.0723 Volume Ratio i

Purification Flow (gpm) 72 72 Purification Flow, yearly average for 0.48

.48 boron control (gpm)

-5

-5 Boron Concentration Reduction Rate (ppm /second) 4.58x10 4.58x10 Beginnings of Life Boron Concentration (ppm) 1200 1200 E

lon Exchanger and Gas ' Stripper removal efficiency CVCS Purification Ion Exchanger Noble gas, tritium 0

0 Cs, Rb

.5

.5 All others

.9

.9 CVCS Lithium Removal Ion Exchanger (2)

("/

Noble gas, tritium 0

0 E

All others

.9

.9 Amendment E December 30, 1988

CESSAR EnnflCATl3N O

TABLE 11.1.1-1 (Cont'd)

(Sheet 2 of 2)

BASIS FOR REACTOR COOLANT FISSION PRODUCT ACTIVITIES PARAMETER MAXIMUM PERFORMANCE (I)

CVCS Gas Stripper Removal Efficiency Noble gas

.999

.999 All others 0

0 CVCS Gas Stripper Operation Continuous None fissignProductEscapeRateCoefficients (sec )

-8

-8 Noble Gases 6.5 x 10 6.5 X 10

-8

-8 Halogens 1.3 x 10 1.3 X 10

-8

~0 CS 1.3 x 10 1.3 X 10 Te, Mo 1.0 x 10 1.0 X 10~9

~9 All Others 1.6 x 10-12 1.6 X 10 -12 N6fts:

(1)

Conditions for use in system operational performance evaluations.

(2) Nuclides are also removed from the letdown flow via the CVCS Lithium Removal lon Exchanger.

This ion exchanger is used in series with the CVCS Purification lon Exchanger during approximately 20% of the core cycle.

E O

Amendment E December 30, 1988

CESSAR 88WnCADON O

TABLE 11.1.1-2 MAXIMUM ACTIVITIES IN THE REACIO.R C00LAf1T DUE TO CONTINUOUS OPERATION AT MAXIMUM POWER WITH_.ONE-0VARTER PERCENT FAILED FUEL E

E (LATER)

O O

Amendment E December 30, 1988

CESSAR inlincum.

O TABLE 11.1.1-3 REACTOR COOLANT SYSTEM ACTIVITIES FOR SYSTEM PERFORMANCE EVALUATIONS E

(LATER)

\\

i I

O Amendment E December 30, 1988

CESSAR EMMnc m.

l I

O v

l 11.1.2 DEPOSITED CRUD ACTIVITIES The activity of radioactive crud and its. thickness on primary system surfaces have been evaluated using measured data from j

various operating pressurized water reactors.

Even though these reactors have different water chemistries and different materials in contact with the primary coolant, their crud activity (dpm/mg-crud), crud film thicknesses and dose rates l

due to this crud are remarkably similar.

The half-lives, reactions and gamma-decay energies for each of the long-lived isotopes in the radioactive crud are as shown in Table 11.1.2-1.

I The long-lived isotopes are those significant isotopes remaining after 48-hours decay.

The radioactive crud originates on in-core and out-of-core surfaces.

The crud plates out on the in-core surfaces and reerodes after a short irradiation period.

This irradiation period or core residence time for each isotope is determined by the following equations (see Appendix 11-A for the derivation of these equations):

A.

Circulating Crud:

))

A A

16.67 g

T In (1 -

),

secs (1) tres "

A E

A i

c l

1 I

B.

Deposited Crud:

A 16.67 y

q res "

5 1" (1 -

), secs (2) t Ifp A, A are the crud activities for each isotope Where:

y g,

(dpm/mg-crud),

2 A

is the total primary system area (cm ),

T Eg4 is the activation rate (d/g-sec), and 2

A is tha core surface area (cm ),

c The' activation cross-section E is as follows:

i (a/o)g (w/o)y Ng oi 2

Og mM (3)

Ei=

[Aji 11.1-5

a CESSAR En9icari:n O'

Where:

(a/o)1 is the isotopic abundance, i

(w/o)y is the elemental abundance in the crud l

or the elemental abundance in the base j

metal, l

24 N

is Avagadros number (0.6023 x

10 a/g-mole),

)

l

[A]1 is the atomic weight of isotope (i), and I

is the microscopic cross-section cr i (barns).

1 1

Circulating crud is taken to be all crud in the reactor coolant.

Deposited crud is taken to be all crud which plates out on in-core surfaces.

i The measured average and maximum crud activities (dpm/mg-crud) as taken from References 7-20 for those reactors considered in the determination of the core residence times are as shown in Table 11.1.2-2.

The average and maximum core residence times as determined by the above expressions, the activities in Table 11.1.2-3 and the system parameters in Table 11.1.2-4 are as shown in Table 11.1.2-5.

As all the Fe-59 residence times are long, its activity (A ) is assumed saturated.

The averages (T of themaximumresfdencetimesarealsogiveninTable11.1.2Fs) 5 l

The calculated crud activities (A,) are determined utilizing the averages (T of the maximum core residence times, the systems parameters [88) Table 11.1.2-6 and the following equation:

& (1 - e es)

(0.06), dpm/mg-crud (4)

Ay=Z1 T

As the averages (T of the maximum residence times are used and in general thedes()T are a factor of 2 to 4 higher than a residM8e) time, the resulting calculated crud straight average activities will be conservative.

These calculated crud activities of the long-lived isotopes are as shown in Table 11.1.2-7.

These calculated crud activities are applied to both the circulating crud and out of core deposited crud.

Using the average crud level in the reactor coolant (75 ppb) of those operating reactors shown in Table 11.1.2-2 and the calculated crud activities (dpm/mg-crud) as shown in Table 11.1.2-7, the i

l average isotopic activities in the primary coolant are determined by the following expression:

O 11.1-6

1 1

l CESSAR innnc.m.

'N j

l l

3 A=

(75 x 10~ ) p (2.7 x 10-5) 1 x 10+3 ci/cm (5) where p is density of water (g/cc) and 1000 is mg/g.

-1 The average calculated activities in the primary coolant using i

the above expression are shown in Table.11.1. 2-8.

The maximum coolant activities can be higher due to " crud bursts" during l

shutdowns or changes in power.

However, these " bursts" occur over short periods of time, and therefore, the average values are more reasonable to use for long term operation.

2 The equilibrium thickness of radioactive crud film (mg-crud /cm )

has been detcrmined by two methods:

A.

The direct measurement of the film during maintenance and/or tests in operating reactors.

B.

Calculating crud film thickness from measured dose rates and specific activities (dpm/mg-crud) of deposited crud.

The equilibrium crud film thicknesses for various Reactor Coolant System areas are as shown in Table 11.1.2-9.

The calculated crud activities in this section are reasonable values and together with measured plateout thicknesses match measured shutdown dose rates around various equipment associated j

with operating reactors.

However, both the crud levels and plateout thicknesses do have rather wide variations as shown in Table 11.1.2-2 for operating reactors and many combinations of activities and plateout thicknesses could reproduce the measured shutdown dose rates, j

The conservative evaluation of the above operating data yields circulating crud concentrations as per Table 11.1.2-8.

O t

Amendment E 11.1-7 December 30, 1988

CESSAR MANICATI@N l

THIS PAGE INTENTIONALLY BLANK O.

O Amendment E 11.1-8 December 30, 1988

CESSAR !!ninCAMN O

i TABLE 11.1.2-1 LONG-LIVED ISOTOPES IN CRUD Isotope h

A.d-I Parent Reaction d is E(mev) 60 5.26Y 3.6(-4) 59 n,y 2.00 1.25 C0 00 58 71.4d 9.73(-3) 58 n,p 1.00 0.81 Co Ni 54 313d 2.21(-3) 54 n,p 1.00 0.84 Mn Fe 51 27.8d 2.49(-2) 50 n,y 0.10 0.32 Cr Cr 59 45d 1.54(-2) 58 n,y 1.00 1.18 Fe Fe 95 65.5d 1.06(-2) 94 n,y 2.00 0.75 Zr Zr O

1' 1

1

3 13 5

f 1

11 1

6 7

e 7

8 1

1 R

02 4

9 11 1

O b

50 00 0 -

50 5 -

7 -

1 80 90 7

55 7

7 4

o 0

4 2

4

)

)0 u

91 C

++

4

((

6 12 31

))

56

))

66 r

++

55 55

+ +

Z

((

++

((

5

((

96 9

68 07 32 22 61 79 24 1 7

)

dur

))

)

c f

65 5

H

++

+ -

a 1

((

(

m 8

26 5

/

1 m

57 1 -

o d(

)

Y

))

6 T

e 67 56 67 66 66 66

+

I F

+ +

++

+ +

++

+ +

(

2

)

V 9

((

- 3 2

I 5

85 78 58 20 88 23 2

T 0

f C

21 63 51 16 1 1 23 2

1 o

A 1

1 D

OL 1

U

)

t R

))

7 E

e C

r 77 67 78 78 66 66

+

e C

+ ^

++

+ +

(

B h

E 1

((

4 A

S V

5 36 10 74 01 22 72 2

T

(

I T

13 32 11 91 22 58 4

CA0 I

)

D

))

5 A

n 67 66 67 67 66 56

+

R M

+ +

++

+ +

++

(

4

((

- 0 D

5 32 42 59 94 37 70 E

7 R

21 14 41 31 1 1 72 9

U S

AE

)

M

))

7 o

78 78 78 78 66 66

+

C

++

+ +

(

8

((

- 3 5

90 22 32 75 82

6. i 5

94 21 31 21 23 49 4

)c

(

)

o

))

6 C

67 67 67 67 77 66

+

0

++

+ +

++

(

6

((

- 2 15 00

7. l 32 38 89 2

92 22 62 42 24 12 2

)

ex ex

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CESSAR innncmo.

O TABLE 11.1.2-6 ASSUMED SYSTEM PARAMETERS. SYSTEM 80+

Parameter Value*

2 Thermal Flux (n/cm -sec) 6.14(+13)

E 2

Fast Flux (n/cm -sec) 2.92(+14)

A /A (LATER)

T c Assumed Activation Rates 3817 Mwt Plant 1

/9 Isotope Activation Rate. E;d>(d/a-sec)*

1 V 60 (LATER)

C0 58 (LATER)

C0 54 (LATER)

Mn SI (LATER) 1 Cr 59 (LATER) 7g I

95 (LATER)

Zr

Number in parentheses denotes power of ten.

i Amendment E December 30, 1988

CESSAR !!!nne.m.

O TABLE 11.1.2-7 LONG-LIVED CRUD ACTly_ITY FOR A STANDARD 3817 MWT PLANT Isotope Tres. (days)

Hal f-Li fe Act, dom /mq 60 166 5.26 y (LATER)

C0 58 22 71.4 d (LATER)

C0 54 110 313 d (LATER)

Mn SI 12 27.8 d (LATER)

E Cr 159 Sat.

45 d (LATER)

FE 95 29 65.5 d (LATER)

ZR O

O 1

Amendment E December 30, 1988

I CESSAR !!!Mnc n:.

TABLE 11.1.2-8 AVERAGE CALCULATED REACTOR COOLANT CRUD ACTIVITY "

l Isotope Act, (uci/cc) 60

(

)

Co 58

(

)

Co 54

(

)

Mn 51

(

R)

Cr E

59 (LATER)

Fe 95

(

R)

ZR

\\

l l

NOTES:

(a)

Reactor coolant temperature is 70'F.

Crud level 75 ppb.

l E

Amendment E December

.";0, 1988

CESSAREnnncm.

TABLE 11.1.2-9 EQUILIBRIUM CRUD FILM THICKNESS Thickngss Location (ma/cm )*

Vessel Internals, Piping 1.00(+0)

SG Inlet Plenum l

Pressurizer Lower Head 6.5(-1)

Surge Line 1.20(+0)

CRDM, Vessel Head ICI Tops 3.00(-1)

SG Tubing 1.00(-1)

Regenerative HX 3.50(-1)

]

Letdown HX 3.00(-2)

Shutdown Cooling HX 3.00(-2)

Number in parentheses denotes' power of ten.

CESSAR n.iWicarien 1

p 1

-(

1 I

11.1.3 TRITIUM PRODUCTION IN REACTOR COOLANT l

The principal sources of tritium production in a pressurized l

water reactor (PWR) are from ternary fission and. neutron induced l

reactions in boron, lithium and deuterium that are present in the coolant, borated shim rods and control element assemblies (CEA).

The tritium produced in the coolant contributes immediately to the overall tritium activity while the tritium produced by fission and neutron capture in,the'CEA's and borated shim rods contributes to the overall tritium activity via release through the cladding.

)

1 11.1.3.1 Activation Sources of Tritium 1

The activation reactions producing tritium are as shown in Table 11.1.3-1.

The tritium production from reactions'5 and 6 (B-11 l

and N-14 sources) is insignificant due to low cross section j

and/or abundance and can be neglected.

Reactions 1-4 (from B-10, j

lithium, and deuterium) are'the major sources of tritium in the 1

coolant, CEA's and borated shim rods.

The tritium production from the above sources is determined by the following expressions:

(

Tritium Formation Rate = Production Rate - Decay

=E4~

a 3

~

N=

(1 - e

),

atoms /cm at time (t)

-11 activity (curies) = VAN x 2,7 x 10

-11

=E4 (1 - e

) V x 2.7 x 10 Where:

Ed is the production rate (atoms /cc-sec) a t

is the reactor operating period of interest i

V is the effective core borated shim rg volup)e,

L volume or CEA volume (cm and 2.7 x

10 l

converts disintegrations /sec to curies.

The appropriate parameters used in the calculation are as shown in Table 11.1.3-2.

Based on these parameters, the tritium O

produced from activation sources in the reactor coolant for one equilibrium fuel cycle are included in Table 11.1.3-3.

11.1-9

CESSARnniNua 11.1.3.2 Tritium From Fission The ternary fission production of tritium in the core is expressed simply by:

= YF - AN N=f(1-e

), atoms at time (t)

~

activity (curies) = AN x 2.7 x 10

-11

= YF (1 - e-At) x 2.7 x 10 Where Y

is the tritium fission yield F

is the fission rate (f/sec) t is the reactor operating period of interest, and

-11 2.7 x 10 converts disintegrations /sec to curies.

Tritium as a pr duct of fission (References 21, 22) has a yieg 5

of 8.0 x 10 atoms / fission for U-235 and a yield of 2.6 x 10 atoms / fission for U-238, Pu-239 and Pu-241.

The amount of tritiNm that is released through fuel cladding can be indirectly determined using measured tritium levels from operating PWR's, subtracting the calculated tritium activity produced by neutron capture in the reactor coolant, and attributing the remaining tritium activity to release from the cladding of the fuel rods, borated shim rods and CEAs.

Due to the large number of fuel rods as compared to the number of borated shim rods and CEAs within i

the core during operation, any amount of tritium released to the system will be principally from the fuel rods.

The total amount of tritium produced per fuel cycle can be determined by summing the total tritium discharged in the gaseous, liquid and solid waste discharges of the plant and the tritium inventories in the Reactor Coolant System and other waste n refueling tanks that can contain tritium at the end of the fuel cycle of interest.

This method has been used to analyze C-E operating data and data from other PWRs (References 25 through 30).

The results of the analysis are shown in Table 11.1.3-4.

Buildup of plutonium in the fuel with burnup was accounted for in the analysis.

Based on this data, an average expected tritium release from the fuel of 2% and a maximum design value of 5% are used to estimate the annual tritium production in Table 11.1.3-3.

E O

Amendment E 11.1-10 December 30, 1988

C E S S A R in nnca m.

O TABLE 11.1.3-1 TRITIUM ACTIVATION REACTIONS Reaction Threshold Enerav (MeV)

Cross Section (mb)(a) 1)

8 (n, 2a)T 1.9 1.13(+1)(b) 10 7

2)

Li (n, na)T 3.9 9.80 6

3)

Li (n, a)T Thermal 9.45(+2) barns 4)

D (n, a)T Thermal 5.70(-1) 5)

B (n, T)9Be 10.4 8.00(-3)

II 6)

N (n, T)12C 4.3 3.00(-1)

I4 O

NOTES:

(a) Threshold cross sections are from References 3 and 4.

(b) Number in parentheses denotes power of ten.

O

- - v-hh bar $1 CATION 1

U(~h TABLE 11.1.3-2 PARAMETERS WHICH INFLUENCE TRITIUM PRODUCTION DETERMINATION 2.47(+7)I )

Effective Core Volume, cm Average Thermal Fission Rate, f/Mw-sec 3.10(+16)

Lithium Concentration, ppm.

Average.

1.3 E'

Maximum 2.3 Lithium-6 Abundance, %

1.6 Boron Concentration, ppm UO 600 2

Power Level, Mwt

)

Average 3800 Maximum 4100 i

Fuel Release, %

Average 2

Maximum 5

NOTE:

(a)

Number in parentheses denotes power of ten.

Amendment E December 30, 1988

CESSAR !!Minc4Tien TABLE 11.1.3-3 CONSERVATIVE ESTIMATE FOR TRITIUM PRODUCTION IN REACTOR COOLANT (Ci/yet,r)

AVERAGE Reaction 2

D (n,y)T 10 6Li (n,a)T 1392 Li (n, na)T 8

OB (n, 2a)T 788 Fission 636

(

Total 2834 MAXIMUM D (n,y)T 10 Li (n,a)T 2320 Li (n, na)T 13 B(N, 2a)T 788 Fission 1336 Total 4467 Amenument E December 30, 1988

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CESSARHiMem,.

Og 11.1.4 NEUTRON ACTIVATION PRODUCTS 11.1.4.1 Nitrocen-16 Activity Nitrogen-16 is' produced by the O(n,p)16N reaction.

Nitrogen-16 G

decays by beta emission and high energy gamma emission 78% of-the time.

The gamma energies are 6.13 mev, 73% of the time and 7.10 mev, 5% of the time.

The nitrogen-16 half' life is '7.13 seconds.

The threshold energy for the reaction is 10.2 mev.

The nitroggn-16 activity at thy pressure vessel outlet nozzle is l

I 5.76 x 10 disintegrations /cm -sec.

This activity is based on the following expression and reactor parameters.

3 Activity (disintegrations /cm - sec) -

(1 - e t) l 3

Where:

E&

is the reaction rate (4.72 x 10 d/cm

- sec),

t is the core transit time (0.79 sec),

c t

is the total primary loop time (8.6 sec),

t v

t is the time from the active core outlet to the point of interest (0.69 see to outlet nozzle) and A

is the decay constant (0.097 sec~ )

11.1.4.2 Carbon-14 Production 17 Carbon-14 is produced in the RCS by igctivation of 0 and N 7

is g roduced by the O isotopes.14 The greatest amount of C a lesser amount of C yps produced by the (g a) C

gaction, N

(n, p) C reaction.

The production of C from both sources can be calculated by using the following equation:

Q=No mtps g g Where:

N

= atom concentration in the RCS water, (atoms /kg H O) 2

= thermal cross section (cm )

o g 13 2

= thermal neutron flux, 5.5 x 10 n/cm -s 4

m

= mass of core water, 2.47 x 10 p

11.1-11 1

CESSAR E!L"icari:n l

I O

t

= conversion factor (sec/yr) p

= plant capacity factor, 0.8

}

~

s

= 1.03 x 10 Ci/ atom Q

= production rate, Ci/ year 17 1.3 x 10 atoms C-14 production from 0 acti_gtion,, No Fg/kg

=

2.4 x 10 cm" are used in the above (H O) and a

=

O equation. 2 The pr8 duction rate is 11.0 curies / yea 50 For 14 bon-14 production from N act_igtiop No = 2.75 x 10 atoms cg/kg 1.8 x 10 cm are used in the above (H O) and a

=

Nequation. 2 The prodtiction rate is 1.8 curies / year.

14 The annual production of C from these sources will be 12.8 curies.

11.1.5 FUEL EXPERIENCE Fuel experience is discussed in Section 4.2.3.2.10.

On the basis of experience accumulated to date, it is expected that the failed fuel fraction during normal operation will be less than 0.12 percent.

11.1.6 LEAKAGE SOURCES Systems containing radioactive liquids are potential sources of leakage to the environment.

Leakage to the containment from the RCS is expected to be at a rate that will allow 1%/ day of the H

primary coolant noble gas inventory and 0.001%/ day of the primary coolant iodine inventory to enter the building atmosphere.

Leakage from all systems located in the auxiliary building is expected to be less than 160 pounds per day.

This leakage is made up from such potential sources as pump gland seals and valve packing.

Primary to secondary steam generator leakage is expected to be less than 100 pounds per day under normal conditions.

Table 11.1.6-1 provides a listing of leakage values from valves and pumps.

Releases inside the plant are handled by the appropriate ventilation system.

l E

Means of controlling leakage are discussed in Section 5.2.

O' Amendment E I

11.1-12 December 30, 1988 i

CESSAR !!ninc m.,

i TABJE__31.1. 6-1 LEAKAGE ADBUMPTIONS FROM C-E SUPPLIED EOUIPMENT Valves Disk Leakage 10 cc/hr/ inch Seat Diameter Stem Leakage 10 cc/hr/ inch Stem Diameter Pumns Centrifugal 50 cc/hr Positive Displaceraent 1 gallon /hr Flanges 30 cc/hr H

i

CESSAR n!Wicma C

11.1.7 SPENT FUEL POOL FISSION PRODUCT AND CORROSION PRODUCT ACTIVITIES corrosion lE Spent fuel pool maximum and expected fission and of the product specific activities are evaluated for the start refueling period.

It. is assumed that upon shutdown for refueling, the Reactor Coolant System is cooled down for a period of approximately two days.

During this period, the primary' coolant is letdown through the purification filter, purification ion exchanger, gas stripper and volume control tank.

This serves two purposes; removing the noble gases in the gas stripper avoids' large activity releases to the containment following reactor vessel head removal, and the ion exchange and filtration. reduces dissolved fission and corrosion products in the coolant which would other-wise enter the spent fuel pool and refueling ' water cavity.

At the end of this-period, the coolant above the reactor vessel flange is partially drained.

The reactor vessel head is unbolted and the refueling water cavity is filled from the in-containment refueling water tank.

The remaining reactor coolant volume is.then mixed with water in. the refueling cavity from the spent fuel pool.

After refueling, the spent fuel pool E-is isolated and the water in the refueling cavity is returned to the in-containment refueling water tank.

This series of events

(~'

determines the total activity in the spent fuel pool.

The spent fuel pool activities will be subsequently reduced by decay during refueling as well as by operation of the Fuel Pool System.

There is no contribution from defective fuel elements because of low power and temperature during storage and " degassing" during plant shutdown operations.

E 11.1.0 STEAM GENERATOR ACTIVITY MODEL The specific activities in the steam generator and secondary systems are based on the parameters supplied in Table 11.1 8 1.

E i

l Amendment E 11.1-13 December 30, 1988

CESSAR !!nincaic, 1

0\\

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)

THIS PAGE INTENTIONALLY BLANK 1

l i

i O

l 0

Amendment E 11.1-14 December 30, 1988

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CESSAR !!nincam, o

TABLE 11.1.8-1 BASIS FOR STEAM GENERATOR LIOUID ACTIVITIES Parameter Averaae Maximum Reactor Coolant Activities Table 11.1.1-3 Table 11.1.1-2 Primary to Secondary Leak Rate, pounds / day 100 12,000 Main Steaming Rate (MSR), pounds / hour *-

Table 10.1-1 Same Steam Generator Blowdown Rate Table 10.1-1 Same E

1 Steam Generator Liquid Mass, lbm (100%)

280,982 Same Steam Generator Partition Factors **

Tritium, Noble Gases 1.0 lodines 0.01 All Other Nuclides 0.001

\\

Values are for sum of both steam generators.

l

- Ratio of radioisotope concentration in gaseous phase to. liquid phase at equilibrium conditions.

i i

O Amendment E December 30, 1988

CESSAR8Binem.

l

\\ })

/

l

11. L i RADIOACTIVE WASTE SYSTEMS information regarding fluid system interfaces with the radioactive waste systems is provided below.

A.

Liquid Waste Management System (LWMS) 1.

Chemical addition package strainer drain Chemical Nature Primary grade water and 2060 ppm LiOH(ax) 2.

Supply to LWMS waste condensate tank E

chemical Nature Primary grade makeup water j

l l

3.

Supply to LWMS waste concentrator l

Chemical Nature Primary water 4.

BAC Drains f-s Chemical Nature Primary water and component

(

cooling water B.

Solid Waste Management System (SWMS) 1.

Ion exchanger resin sluicing lines Chemical Nature Resin, air, reactor makeup water 3 Volume of dewatered 36 ft / ion exchanger sluicing i

resin operatjon i

Volume of resin 180 ft discharged per year (based on one replace-j ment per resin bed per j

year) 3 Spent resin design Table 11.1.9-1 l

activity input to SWMS i

2.

Strainer blowdown lines j

Chemical Nature Resin slurry j

E 3.

Boric acid concentrator concentrate discharge to SWMS 1

Chemical Nature 12 wt % boric acid (max) l fs

(

)

Volume 2000 gallons (max) l

\\s /

Amendment E 11.1-15 December 30, 1988

CESSAR 8a'JPICATCN O

4.

The Solid Waste Management System shall be capable of receiving the following quantities of spent filter cartridges or equivalent each year:

Replacement Frequency Waste Volume j

j Seal injection filters 2

4.18 ft 3

Purification filters 4

16.72 fg Boric acid filters 1

2.09 ft Reactor drain filters 1

2.09 ft 3 Reactor makeup filters 1

2.09 ft C.

Gas Waste Management System (GWMS)*

1.

Purification and deborating ion exchanger vent Chemical Nature Air Volume 170 scf/ year E

2.

In-Containment Refueling Water Tank vent Chemical Nature Air Volume 86,000 scf/ year 3.

Volume Control Tank Gas relief Chemical Nature H

or N Volume 1300 sch/ year 4.

Volume Control Tank vent Chemical Nature H

and/or N l

Volunie 3kl0 scf/ye$r I

i 5.

Volume control Tank gas sample line Chemical Nature H, and/or N 2 Volume Small 6.

Reactor Drain Tank vent Chemical Nature N,H Volume 7h59 scf/ year i

The design concentrations of radionuclides in the principal waste stream inputs to the GWMS are provided in Table 11.1.9-2.

l Amendment E 11.1-16 December 30, 1988

CESSAR EDUricari:n

{

/O

\\~-)

7.

Preholdup lon exchanger vent I

Chemical Nature Air Volume 60 scf/ year 8.

Equipment Drain Tank vent Chemical Nature N

Volume 7359 scf/ year 9.

Equipment Drain Tank gas sample line Chemical Nature N

Volume Sball 10.

Holdup Tank vent Chemical Nature Air E

Volume 160,000 scf/ year 11.

Holdup Tank gas to analyzer Chemical Nature Air I

Volume Small

(

j 12.

Boric Acid Concentrator vent Chemical Nature Air, N, water vapor Volume 1400 sbf/ year 13.

Boric Acid Condensate ion exchanger vent Chemical Nature Air Volume 100 scf/ year 14.

Reactor Makeup Water Tank vent Chemical Nature Primary grade water Volume 67,920 scf/ year 15.

Gas Stripper Vent to gas surge header Chemical Nature H

noble gases Volume 1k5,672 scf/ year 16.

Gas Stripper Gas to analyzer Chemical Nature H,

noble gases i

Volume Sball

\\w Amendment E 11.1-17 December 30, 1988 l

l CESSAR !!ai"1CATICN O\\

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l i

I O

Amendment E 11.1-18 December 30, 1988

CESSAR !!nincuie.

O TABLE 11.1.9-1 ANNUAL SPENT RESIN ACTIVITY INPUT TO SWMS (Curies)~

(LATER)

E O

j O

Amendment E December 30, 1988

CESSAR EMMncano i

f l

TABLE 11.1.9-2 SPECIFIC ACTIVITIES OF SOURCES TO THE GWMS l

DURING NORMAL. OPERATION I

(pci/cc)

]

l I

(LATER)

I E

l 1

l l

i l

i

\\

Amendment E December 30, 1988

~

J CESSA.R !!nLm.

l

(

l REFERENCES FOR SECTION 11.1

]

l 1.

M.

E.

Meek, B.

F.

Rider,

" Summary of Fission Product i

~!

Yields", NED012154, January 1972.

2.

" Chart of Nuclides," USAEC, Modified by Battelle-Northwest, May 1969 and May 1970.

1 3.

" Neutron Cross Sections,"

BNL 325 Supplement No.

2,,

May 1964.

q 4.

J.

D.

Eichenberg,

" Effects of Irradiation on Bulk UO 2'

WAPD-183, October 1957.

j 5.

G.

M.

Allison and H.

K.

Rae, "The Release of Fission Gases i

and Iodines from Defected UO Fuel Elements of Different Lengths," AECL-2206, June 1965 6.

Deleted.

E' 7.

Connecticut Yankee Monthly Operating Reports, 2/68, 3/68, 6/68, 7/68, 12/68, 1/69, 3/69-5/69, 8/69, 10/69, 12/69, 3/70, 10/70, 11/70.

8.

San Onofre Monthly Operating Reports, 1/71-3/71, 6/71-9/71, 11/71, 12/69, 1/70.

9.

Yankee Rowe Monthly Operating

Reports, 2/69-6/69, 8/69-12/69, 1/7012/70, 1/72, 4/72-7/72.

10.

Large Closed-Cycle Water Reactor Research and Development Program, Progress Report January 1, 1965 March 31,

1965, WCAP-3620-12.

11.

The Saxton Chemical Shim Experiment, Weisman J.,

Bartnoff S, July 1965, WCAP-3269-24.

12.

Large Closed-Cycle Water Reactor Research and Development

Program, Progress Report April 1, 1965 - June 30,

1965, WCAP-3269-13.

13.

Corrosion Product Behavior in Stainless-Steel-Clad Water Reactor Systems, Nuclear Applications, Vol. I October.1965.

14.

Decontamination of the Shippingport Atomic Power Station, Abrams C.

S.,

Salterelli E.

A., January 1966, WAPD-299.

(

Amendment E 11.1-19 December 30, 1988

l l

l k)! h h k!k bb ' ICATCH l

0 15.

Radiation Buildup on Mechanisms and Thermal

Barriers, Weingart E.,

June 1963, WAPD-PWR-TE-145.

16.

Indian Point 1 Semi-Annual Operations Reports, 9/66, 9/67, 3/68, 9/68.

17.

Test Data Sheets, Maine-Yankee Core Crud Removal, CENPD-113, August 23, 1973.

18.

D.

L. Uhl, Oconee Radiochemistry Survey Program, Semi-annual Report July-December 1973, May 1974.

19.

D.

L. Uhl, Oconee Radiochemistry Survey Program, Semi-annual Report January-June 1974, May 1975.

20.

P.

J.

Grant et.

al.,

Oconee Radiochemistry Survey Program, RDTPL 75-4, May 1975.

l 21.

M E.

Mppk, B2 3h*

359 " ummary pg7 Fission Product Yie ds Fiss} ion q

l for U

U Pu and Pu at

Thermal, Spectrum and 14 MeV Neutron Energies", APED-5398, Class 1,

March 1, 1968.

22.

ANL-7450 Chemical Engineering Division Research Highlights, May 1967 April 1968.

23.

E.

P.

Lippincopy,A.

L.

Pitner and L.

S.

Kellog,

" Measurement of B (n,t) Cross Section in a Fast Neutron 4

Spectrum", HEDL-TME-73-49, May 1973.

l 24.

" Neutron Cross Sections",

BNL 325 Supplement No.

2, May 1964.

25.

Point Beach Semi-Annual Reports, 6/71-1/74.

26.

H.

B. Robinson Semi-Annual Reports, 6/71-1/75.

27.

Ginna Semi-Annual Reports, 6/71-1/75.

28.

Sodrce Term Data for Westinghouse Pressurized Water Reactors, WCAP8253, May 1974.

29.

P.

J.

Grant et.

al.,

"Oconee Radiochemistry Survey Program",

RDTPL-754, May 1975.

30.

Omaha Semi-Annual Reports, 1973-1975.

O 11.1-20 l

l

CESSAR E! Enc m.

A k

i v

)

11.2 LIOUID WASTE MANAGEMENT SYSTEM 11.2.1 DESIGN BASES 11.2.1.1 Criteria The Liquid Waste Management Systen (LWMS) is designed to meet the following criteria.

A.

The system must meet the regulatory design basis which is that it be capable of reducing releases of radioactive material in liquid effluents to as low as reasonably achievable in accordance with 10 CFR 50 Appendix I.

B.

The system must contribute to meeting the performance design objectives in that it must never interfere with normal station operation including anticipated operational occurrences.

C.

The system must meet the safety design basis which is that the consequences of accidental releases from the LWMS must not exceed the Standards for Protection Against Radiation, E

10 CFR 20.

V D.

The system must also contribute to meeting the occupational exposure design objective by keeping operation and maintenance exposure as low as reasonably achievable.

11.2.1.2 Codes and Standards The LWMS is designed in accordance with the guidance of Regulatory Guide 1.143 and is designed to the codes and standards listed in Table 1 of Regulatory Guide 1.143.

Although the LWMS is not required to be designed as Seismic l

Category I, it is surrounded by a curb, capable of retaining the entire liquid contents of the Radwaste Building.

The foundation and surrounding curb are designed to withstand or accommodate long term settlement.

11.2.1.3 Features The following features assist in meeting the Design Criteria:

A.

The system takes advantage of feed stream segregation to allow most efficient processing of waste.

This permits large volumes of potential waste with little or no contamination to be monitored and released.

It also O

prevents small volumes with high chemical or radiological Q

impurity from controlling the entire process.

Amendment E 11.2-1 December 30, 1988

CESSAR 8HMcmou O

B.

The treatment processes used are as simple as possible.

The primary process is ion-exchange which can be specifically altered for particular waste.

C.

The system has provisions to accommodate leased equipment which may provide the most economical processing at particular times or for particular waste.

D.

Many normal system operations are remote from a centralized control panel which permits operators to most effectively coordinate activities.

E.

Active and replaceable components have crane access to facilitate removal and repair.

F.

The system is capable of accommodating new configurations and processes as they may become available.

The system is arranged so that these changes can be made with minimum cost and time.

11.2.2 SYSTEM DESCRIPTION E

11.2.2.1 General Description The LWMS consists of collection tanks, process pumps and vessels, sample tanks, and appropriate instruments and controls to permit most operation to be conducted remotely.

The process equipment and connection for mobile leased equipment are located so that they have crane access for ease of repair, replacement, or reconfiguring.

The principal process is ion exchange.

This allows flexibility for tailoring the process to naecific wastes while retaining simplicity of operation.

The precess vessels can also be loaded with granulated carbon or other sluicable media which may be advantageous for filtration or adsorption.

Remote valves and piping are arranged so that process vessels may be used cyclically on a given waste stream.

That is, the newest vessel may be placed last in the stream and moved closer to the head of the stream as it becomes loaded.

Collection tanks and sample tanks are provided in pairs so that one may be available to receive waste while the other is being processed or discharged.

Subsystems are provided for non-recyclable equipment

drains, floor drains, and orgonic wastes.

Collection / Holding tanks are also provided for chemical wastes and oil separators.

These O Amendment E 11.2-2 December 30, 1988

CESSAR EnWnc41,2n 1

,0 wastes are either-reprocessed in the collection / holding tank and then treated as floor drains or directly packaged / solidified for shipment.

11.2.2.2 Components Description 11.2.2.2.1 Waste Collection Tanks The Equipment Drain, Floor Drain, and Laundry Collection Tanks (2 cach) are each sized for the anticipated peak daily input taking into account anticipated operational occurrences but not considering events which might occur less often than once per fuel cycle.

The Waste Collection Tanks are all equipped with mixers, manways and material addition ports accessed from the top of the tank.

The tanks are all located in lined rooms in which the walls E

constitute the appropriate shielding.

11.2.2.2.2 Waste Sample Tanks The Equipment Drain, Floor Drain, and Laundry Sample Tanks (2 each) are each sized for the anticipated peak daily input taking l

into account anticipated operational occurrences but not considering events which might occur less often than once per fuel cycle.

The Waste Sample Tanks are equipped with mixers designed to produce uniform contents for sampling and manways accessed from the top of the tank.

The Waste Sample Tanks are all made of stainless steel and are designed for atmospheric pressure plus friction loss in the overflow lines.

11.2.2.2.3 Process Pumps Each waste stream is provided with a centrifugal pump which can be cross connected with another in case of failure.

Pumps can be flushed and drained prior to maintenance activity and can be readily replaced with on site spares if necessary.

The wetted parts of the pumps are corrosion resistant in order to minimize the buildup of contamination and prolong their life.

11.2.2.2.4 Process Vessels The process vessels are stainless steel pressure vessels with inlet distributors, screened outlets and sluice outlets.

The 7

g normal use of the process vessels is as an ion-exchange bed and s

Amendment E 11.'2-3 December 30, 1988

CESSAR EL"icari:u O

an incidental filter.

Other sluicable beds such granular carbon may also be loaded in addition to or instead of an ion-exchange f

resin.

]

I Access is provided to manually load the vessels if appropriate.

i The normal disposition of a fully expended (high differential pressure, high radiation or loss of removal capability) bed is to be sluiced to a High Integrity Container in the Solid Waste Management System (SWMS).

11.2.2.2.5 Provisions for Mobile Equipment It is anticipated that it may be advantageous to use mobile treatment or direct solidification equipment at times.

This may be true because of changing waste streams or changing economics of processing, shipping and burial.

Piping provisions are made to permit connection of mobile process equipment while using the installed Waste Collection Tanks and Waste Sample Tanks.

11.2.2.2.6 Steam Generator Drain Tank A Steam Generator Drain Tank is provided so that in the event of a steam generator tube leak, the affected steam generator can be drained after isolation.

This water will be generally unsuitable r

for recycling because it will be chemically unsuitable for the Reactor Coolant System and radioactively unsuitable for the Condensate system.

The Steam Generator Drain Tank is sized for three Steam Generator volumes of feedwater to allow for a rinse.

The tank is made of Stainless Steel to permit it to be used as a general waste overflow storage volume if necessary; however, it is normally kept empty.

11.2.2.3 System Operation During normal operation, each pair of Waste Collection Tanks will have one available to accept waste and the other will be available for processing if necessary and discharge.

Since the operators of the LWMS will have level indication on waste volumes at their source, they will anticipato system requirements before they occur.

After a Waste Collection Tank has received as much waste as the operators deem appropriate, its inlet valve is closed to permit sampling, any appropriate chemical addition and processing as necessary, to occur on fixed constituents.

When the collection tank is isolated it will already be essentially mixed because the mixer will have been in operation since the tank exceeded the low level permissive.

A final sample may then be taken with assurance that it is representative.

l I

I Amendment E 11.2-4 December 30, 1988

CESSARnah m e-Based on the sample results, the decision will be made to process the tank using the existing process vessel or to provide a more appropriate process.

Because of segregation of inputs, the size

(

of the collection tanks, and the flexibility of the normal ion-exchange process, a revised process should not normally be necessary.

However, if a change is considered necessary, it will be implemented based on status of individual process vessels inferred from previous influent and effluent sampling.

Re-alignment of the flow path can be rapidly accomplished using remote operated valves.

It is not expected that re-alignment of the process vessels will normally coincide with changes in collection. tanks being

)

processed.

This is because, as indicated above, the variation in I

constituents from tank to tank is not likely to be great, but the l

process beds will routinely become expended.

A bed may be removed from service because of high differential pressure, high radiation or lack of specific removal capacity.

Ideally, a bed i

considered expended for one of those reasons will also be nearing limits for the other as well.

Normally, the first bed in the process flow path will become expended.

When it does, it will be isolated in preparation for transferring the contents to the E

O SWMS, and a new bed will be added at the end of the process flow Q

path.

By this method, each bed will first be the final step of a series of processes and will advance to the first position over its life.

The number of beds in a particular series will be expected to change with circumstances and is left for the operators to determine.

Each pair of sanfple tanks will also alternate as the receiver of the process stream.

The one that is filling will have the mixer started above the low level permissive so that when the tank is l

full, a representative sample will be immediately available.

The details of the effluent release are site specific.

11.2.3 SAFETY EVALUATION The LWMS has no safe shutdown or accident mitigation function.

It is demonstrated in Chapter 15 that accidental release, when evaluated on a conservative basis are not expected to exceed the limits of 10 CFR 20.

11.2.4 INSPECTION AND TESTING REQUIREMENTS A program of testing requirements appropriate to assure that the LWMS is operating as intended is developed prior to fuel loading.

Emphasis is placed on verifying remote

function, and instrumentation important to the design objectives.

(

Amendment E 11.2-5 December 30, 1988

CESSAR nnince O\\

i Testing of the waste streams for the most effective and economical process is required periodically during normal operation.

11.2.5 INSTRUMENTATION REQUIREMENTS I

Instrumentation and indications important to the Design Basis of f

the LWMS are as follows:

A.

Level Indicators All Waste Collection and Waste Sample Tanks are equipped with continuous level indicators.

In addition, redundant means of detecting high level are provided along with j

non-redundant low level indicators.

High level is alarmed both locally and in the LWMS control area.

Levels in the area sumps and tanks which feed the LWMS Collection Tanks are also indicated in the LWMS control area.

B.

Radioactive Liquid Effluent Monitor Prior to release, effluents a held in a sample tank from which a representative sample may be taken.

Inlet valves on E

tanks being released are closed, providing a batch release.

However, all releases are made through an effluent monitor whose set point is adjusted so that it will only alarm on unexpected high activity.

The alarm also terminates the release.

C.

Differential Pressure Each Process Vessel is equipped with dif ferential pressure measurement to monitor the condition of the bed.

D.

Flow 1

Each Process Pump is equipped with flow measurement to assist the operators in regulating the process within appropriate bounds for acceptable effectiveness.

This is also important to judge process bed condition in conjunction with differential pressure.

E.

Area Radiation Area Radiation monitors are discussed in Chapter 12.

O Amendment E 11.2-6 December 30, 1988

CESSAR E50icari:n p

11.3 GASEOUS WASTE MANAGEMENT SYSTEM The design objectives of the Gaseous Waste Management System (GWMS) are to protect the plant personnel, the general public, and the environment by providing means to collect, store, sample, and monitor radioactive gaseous waste.

Design releases of

)

radioactive and potentially radioactive materials to the environment in plant and out are well below concentration specified in 10 CFR 20 and as low as practicable in accordanpe with 10 CFR 50.

11.3.1 DESIGN BASES 11.3.1.1 Criteria The GWMS receives and/or collects waste gases which originate in the reactor coolant system and which require processing by holdup for decay prior to release.

l Effluents normally released to unrestricted areas shall be below the limiting requirements of 10 CFR 20 and meet the "as low as reasonably achievable" (ALARA) objectives of 10 CFR 50, Appendix p

I, and 40 CFR 190.

Control and monitoring of release of radioactive materials to the E

environment shall be in accordance with 10 CFR 50, Appendix A.

l j

The following is included in the control and monitoring release of radioactive materials to the environment:

A.

The gaseous radioactive waste discharges to the environment is monitored by redundant monitors and provided with automatic isolation if the discharge limit is exceeded.

B.

No possibility of gravity or siphon flow from radioactive waste systems to the environment exists.

A 0.25% failed fuel level is design bases for plant gas releases and is met with a minimum dilution.

Accidental release of radioactive material from a

single component of GWMS system does not result in an offsite dose which would exceed the guidelines of 10 CFR 20 (i.e.,

designs shall preclude such releases).

Cost-benefit analyses are made to make releases ALARA.

l l

Amendment E l

11.3-1 December 30, 1988

CESSAREna mu O

11.3.1.2 Codes and Standards The GWMS and associated components are designed in accordance with applicable codes and standards.

11.3.2 SYSTEM DESCRIPTION 11.3.2.1 General Description The GWMS uses charcoal filters to retain the off-gases for the desired time for decay of fission products (xenons, and kryptons) resulting from fuel leakage and tramp uranium on fuel sources.

l The GWMS is designed to receive discharges from the following components; volume control tank, reactor drain tank, refueling failed fuel

detector, and the gas stripper.

The following requirements are met:

l A.

Influent to the gas surge header when collected and held in the GWMS:

Fluid Nitrogen or hydrogen E

B.

Influent to the gas surge header and discharging directly through the discharge header:

Fluid Air, nitrogen, or hydrogen C.

Influent to the mixing header when collected and held in the l

GWMS:

Fluid Hydrogen or nitrogen The following systems only contain radioactive gas when there are concurrent fuel leans and secondary steam generator tube leaks which would transport gases from the primary to the secondary system.

The GWMS does not have the function of processing the gases from the following systems.

A.

Main condenser air ejectors I

l B.

Main condenser mechanical vacuum l

i C.

Turbine gland seal exhaust D.

Deaerator vent O1 l

Amendment E 11.3-2 December 30, 1988

l l

CESSAR55Lmu m

fv Dryers condition the gases to provide the moisture and temperature conditions necessary for the desired performance of

)

the charcoal adsorbers.

The remaining gases are transported to a j

monitored HVAC vent for monitoring and release to the atmosphere.

]

Monitoring and release are functions of the HVAC systems.

1 Holdup for decay of radioactive gases is accomplished by retention in the charcoal beds.

The charcoal beds operate at ambient temperature with the maximum off-gas building temperature defined in accordance with overall plant requirements.

The delay time in the charcoal adsorbers is sufficient to reduce radioactive gas releases to the atmosphere to meet the design l

l objectives of 10 CFR 50, Appendix I, and the limits of 10 CFR 20, while taking into account other potential release sources.

In

addition, sufficient charcoal is included in the charcoal adsorbers to satisfy NRC cost / benefit criteria for reducing environment radiation exposure.

The charcoal mass required is determined using the equation:

M = FT/Kd where l

M E

charcoal' mass, gm

=

flow rate of carrier gas, cc/sec F

=

average delay time, sec T

=

dynamic adsorption coefficient, cc/gm K

=

d All values are those at operating conditions.

The GWMS operates at a

positive pressure relative to the atmospheric to prevent the inleakage of oxygen into the hydrogen carrying piping and the resulting flammable mixture.

A continuous minimum flow of hydrogen (approximately 1 cfm) is maintained in the off gas system.

l l

Where the potential for an explosive mixture of hydrogen and oxygen

exists, the gaseous radioactive waste handling and treatment system is designed either to maintain system integrity by:

A.

Preventing the formation or buildup of explosive mixtures.

l B.

Withstanding the effect of a hydrogen detonation to permit its continued operation.

p l

Amendment E 11.3-3 December 30, 1988

i CESSAR E bmu j

O Where prevention is accomplished, parallel gas analyzers are used to detect the formation of buildup of explosive mixturns, and annunciate both locally and in the main control room for remedial action.

Where appropriate, provisions are included to limit the components / systems that - could have potentially combustible gases enter them.

The off-gas process is continuously maintained non-flammable by maintaining less than 1

percent oxygen in the atmosphere.

l Nitrogen is available for inerting the event of a more than 1 percent oxygen content in the gas stream.

Sample bottles and associated equipment are designed to permit purging with nitrogen prior to use.

j 11.3.2.2 components Description i

11.3.2.2.1 Charcoal Beds The holdup of radioactive gases for decay is accomplished by retention in charcoal beds.

The vessels are designed such that E

if gas flows upward, the bed is not fluidized under any expectr' operating mode.

The vessel is sized such that the superficial velocity through the bed is greater than 1

cm/sec unless j

adjustment in adsorbent volume is made to account for the loss in bed efficiency.

The charcoal beds are designed to allow the loading and unloading of adsorption media.

Operation is at ambient temperature.

A charcoal guard bed is provided upstream of the adsorber beds to protect the beds from contamination and flooding.

The guard bed can be bypassed or purged or reloaded if contaminated.

11.3.2.2.2 Gas Dryers Parallel desiccant dryers are used to achieve proper charcoal bed relative humidity.

Regeneration is accomplished without venting to the atmosphere.

11.3.2.2.3 Piping and Valves Drain lines and valves are sized an continuously sloped to minimize the potential for plugging.

Valves are of the packless metal diaphragm type and have bellows sealed stems to minimize leakage.

All loop seals vent to a controlled vent system and equipment drains are closed or provided with loop seals to limit the escape of radioactive gases.

O Amendment E 11.3-4 December 30, 1988

m l

CESSAREEh m.

V 11.3.3 INSPECTION AND TESTING REQUIREMENTS The GWMS is tested to leak rate limits specified in ANSI /ANS 55.4.

The sum of the leak rates from all' individual components located within a zone does not exceed the zone totals in ANSI /ANS 55.4.

E l

11.3.4 INSTRUMENTATION REQUIREMENTS Table 11.3.4-1 provides a list of instrumentation.

Additionally, hydrogen detectors are provided in compartments containing off-gas rystems under pressure and where hydrogen leakage may occur.

Detection of hydrogen causes the GWMS to automatically shutdown.

Normally, upon reaching a high level setpoint, an alarm annunciated.

Instrumentation in contact with process streams is designed to minimize the potential for explosion.

l Manual ov rride capability of automatic controls is provided for where necessary to maintain system operability.

For the equipment operated manually, remote manual hand switches with status lightc are provided for all frequently operated valves and components.

O i

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l O'O 11.4 BOLID WASTE MANAGEMENT SYSTEM l

l 11.4.1 DESIGN BASES 11.4.1.1 Criteria The Solid Waste Management System (SWMS) is designed to meet the following criteria.

A.

The system must contribute to meeting the performance design objectives in that it must never interfere with normal station operation including anticipated operational occurrences.

B.

The system must also contribute meeting the occupational exposure design objective by keeping operation and maintenance exposure as low as reasonably achievable.

c.

The system must produce-a packaged waste from suitable for shipment to a licensed burial facility.

g l

11.4.1.2 Codes and Standards l

C.

The SWMS is designed in accordance with the guidance of Regulatory Guide 1.143 to the codes and standards listed in Table 1 of Regulatory Guide 1.143.

Although the SWMS is not required to be designed as Seismic Category I, it is surrounded by a curb, capable of retaining-the entire liquid contents of the Radwaste Building.

The foundation and surrounding curb are designed to withstand or. accommodate long term settlement.

11.4.1.3 Features The following features assist in meeting the Design Criteria.

A.

Tha system has provisions to accommodate leased equipment which may provide the most economical choice at particular times or for particular waste.

B.

Many normal system operations are remote from a centralized I

control panel which permits operators to most effectively coordinate activities.

C.

Active and replaceable components have crane access to facilitato removal and repair.

Amendment E 11.4-1 December 30, 1988

1 CESSAREnac-O D.

The system is capable of accommodating new configurations I

and processes as they may become available.

The system is I

arranged so that these changes can be made with minimum cost and time.

11.4.2 SYSTEM DESCRIPTION 11.4.2.1 General Description A primary function of the SWMS is to sluice Process Vessel beds i

from the Liquid Waste Management System (LWMS) to a

High Integrity Container (HIC) and remove the free water from the HIC in preparation for shipment to a licensed burial facility.

The water removed to the HIC is returned to the LWMS and SWMS vents are connected to the Gaseous Waste Management System (GWMS).

The SWMS also includes building space to sort miscellaneous contaminated dry solids and package them appropriately.

A l

compactor is available for dry solids such as paper, plastic and cloth products which can be readily volume reduced.

l Other, more specialized, low volume wastes such as dry cleaning fluid, counter fluid, and contaminated oil will be delivered to an on-site contractor who specializes in solidification / packaging E

of such materials.

Building space and service connections are provided for such contractors.

Finally, provisions are made within the building to interface with a general solidification service contractor should that become desirable or necessary.

4 11.4.2.2 Components Description 11.4.2.2.1 HIC Fill / Dewatering Head HICs are filled using a Fill / Dewatering Head which attaches to HIC and incorporates:

A.

Fill line.

B.

Vent.

C.

Dewatering line inserted in a throw-away, bag

filtered, dewatering standpipe.

The bag filter is sized appropriately to retain the particle size distribution of the

resin, granulated carbon or filter aid material used in the LWMS.

O Amendment E 11.4-2 December 30, 1988

CESSAR 8Hnncm2 4

11.4.2.2.2 Blurry Pump Wet solids are sluiced from the LWMS Process Vessels to the SWM' HICs using an adjustable speed positive displacement slurry pump.

The slurry piping is arranged with a minimum of vertical flow and with cleanout connections and water injection connections.

The Slurry Pump normally takes suction from the dewatering connection j

of the HIC.

There is also a suction connection for sluice water 1

l from the LWMS and a discharge connection to return sluice water to the LWMS.

j t

11.4.2.2.3 Radwaste Building Crane The Radwaste Building Crane covers the areas occupied by the:

(

A.

LWMS Process Vessels B.

LWMS Process Pumps ll C.

Filled HIC Storage D.

HIC Filling Platform j

E.

Shipping Truck Bay V

E F.

Vendor Solidification Bay G.

Miscellaneous Contractor Space H.

Dry Solids Handling Area The Crane is equipped with remote controls and surveillance cameras to minimize operation exposure.

11.4.2.2.4 Dry Solids Compactor The Dry Solids Compactor is used to reduce the volume of such l

material as cloth,

paper, and plastic that is contaminated.

l Sorting and staging space is available in the SWMS area to l

separate non-contaminated materials for ordinary land fill l

disposal.

j

.i 11.4.2.3 System Operation When liquid process vessel beds are

expended, they must be sluiced to an HIC for eventual shipment to a licensed burial facility.

Using Flush water from the LWMS, the Slurry Pump l

injects water into the normal outlet of the expended process l

/n vessel where it flows backwards through the outlet screens.

When i

b' l

the process vessel is filled, the unscreened outlet is opened to 4

Amendment E 11.4-3 December 30, 1988

)

I

i CESSAR8Ebmu j

direct the slurry to the HIC.

After the bed is transferred, the Slurry Pump continues to take suction on the dewatering connection of the HIC until it is suitable for shipment.

l The filled HIC's may be stored in the shielded Filled MIC Storage area for as long as approximately six months prior to shipment.

The Dry Solids Compactor and sorting / packaging operation is completely independent of other Radwaste Systems and is operated whenever the need exists.

j 11.4.3 SAFETY EVALUATION The SWMS has no safe shutdown or accident mitigation function.

It is demonstrated in Chapter 15 that accidental releases, when evaluated on a conservative basis are not expected to exceed the limits of 10 CFR 20.

11.4.4 INSPECTION AND TESTING REQUIREMENTS A Process Control Program appropriate to assure that the SWMS is operating as intended is developed prior to fuel loading.

Emphasis is placed on verifying remote

function, and E

j instrumentation important to the design objectives.

11.4.5 INSTRUMENTATION REQUIREMENTS Instrumentation and indications important to the Design Basis of the SWMS are as follows:

A.

Level Indicators The HIC Fill / Dewatering Head has a high level indicator used to prevent overflow during the sluice operation.

B.

Area Radiation Area Radiation monitors are discussed in Chapter 12.

C.

Flow and Pressure The Slurry Pump has discharge flow and suction and discharge pressure monitored in order to properly control the bed transfer process.

O Amendment E 11.4-4 December 30, 1988

CESSAR CERTIFICATION aa'a" 11.5 PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS (LATER)

E l

'l e;

ent E 11*5-1 December 30, 1988

_

ML20247H353

Text

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