ML17334A486

Forwards Response to NRC 830502 Request for Addl Info Re Assumptions & Methodology Used in Estimating post-accident Radiation Environs for Equipment Items

ML17334A486
Person / Time
Site:	Cook
Issue date:	11/03/1983
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AEP:NRC:0578I, AEP:NRC:578I, IEB-79-01B, IEB-79-1B, NUDOCS 8311110065
Download: ML17334A486 (142)

Text

REGLJLATOWlNFORMATION DISTRIBUTION EM (RIDS)

ACCESSION, -NBR 831 1 1 10065 'OC~DATE ~ 83/1 1/03 NOTARIZED NO DOCKET ¹ FACIL.50 315 Donald C,

Cook Nuclear Power Pl anti Uni t 1i Indiana

'8 0504

-L5 50 31'6iDonal d C, Cook Nuclear Power Pl anti Uni t 2i Indiana 8

AUTH BYNAME AUTHOR AFFILIATION ALEXICH,M~ P ~

Indiana 8 Michigan Electric Co.

RECIP,NAME RECIPIENT AFFILIATION DENTONiH ~ R ~

Office of Nuclear Reactor Regulationi Director SUBJECT>

Forwards response to NRC 830502 r equest for addi info re assumptions L methodology used in estimating post accident radiation environs for equipment items'gt S

z DISTRIBUTION CODE:

A048S COPIES RECEIVED:LTR ENCL SIZE!

TITLE: OR/Licensing Submittal:

Equipment Qualification

. NOTES:

RECIPIENT ID CODE/NAME NRR ORB 1 BC 12 COPIES LTTR ENCL 1

0 RE C'I P I EN T ID CODE/NAME FRIGG INGTONi D 01 COPIES LTTR EiVCL 1

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04 EXTERNAL! ACRS 15 NRC PDR 02 NTIS 1

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TOTAL NUMBER OF COPIES REQUIRED'TTR 26 ENCL 25

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~ g INDIANA8 MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 November 3, 1983 AEP:NRC10578I Donald C. Cook Nuclear Plant Unit Nos.

1 and 2

Docket Nos. 50-315 and 50-316 License Nos.

DPR-58 and DPR-74 EQUIPMENT ENVIRONMENTAL QUALIFICATION PROGRAM1 CALCULATION OF SPECIFIED RADIATION DOSES

" Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

This letter responds to Mr. S. A. Varga's letter dated May 2,

1983, which requested additional information on the environmental qualifica-tion program for electric equipment important to safety at the Donald C.

Cook Nuclear Plant Unit Nos.

1 and 2.

More specifically, Mr. S. A.

Varga requested that we provide a reference for the assumptions and methodology used in estimating post-accident radiation environments for equipment items of concern.

Furthermore, we were requested to include a description of the Donald C. Cook Nuclear Plant containment type and a discussion regarding the, source terms which have been used in the qualification program.

Attachment 1 to this letter provides a general description of the ice condenser reactor containment system utilized by the Donald C. Cook Nuclear Plant.

Please note that the Final Safety Analysis Report for our facility contains additional information which we have not included in this abbreviated description.

Attachment 2 to this letter addresses the basic methodology used in calculating radiation environments both inside and outside the ice condenser containment.

In general, we utilize a computer code which calculates gamma and beta doses due to a cylindrical source which is modeled as a set of line sources.

Versions of this code (e.g.,

NSLSHL3 and SHL1GG) have been used in a number of applications, such as calculating equipment radiation doses for equipment items which are near the recirculation flow path piping systems following a Loss-of-Coolant Accident (LOCA).

We have also applied this methodology to airborne and submerged source term dose calculations for many components inside containment.

8311110065 831103 PDR *DOCK 05000315 P

PDR

Mr. Harold R..

on AEP:NRC:0578I The application of this methodology to equipment items inside containment is described in additional detail in Attachment 3 to this letter.

That attachment provides information on the assumed source term (comprised of fifty-four radioisotopes),

the studies which led to the selection of bounding accumulated dose values as a function of time for components inside containment (both above and below. the containment flood-up level), and sample calculations.

Attachment 4 to this letter presents the assumptions which led to similar bounding accumulated dose values as a function of time for components outside containment.

It is noted that the computer code which was used in the outside containment studies formed the basis for the computer code utilized in our work on equipment within containment.

The methodology employed by each of these codes is, however, similar enough to ensure that no appreciable differences exist between the basic

'alculational,methods used in calculating radiation doses either inside or outside containment.

The results of sample calculations are also included in Attachment 4 for numerous equipment items located near recirculation flow paths in a post-LOCA environment.

It should be noted that we have not enclosed calculations for every equipment item identified in response to either IE Bulletin No.79-01B or 10 CFR 50.49.

Rather, this submittal is intended to provide your staff with an understanding of the basic methodology and assumptions which are used in determining environmental qualification radiation specification doses for many electric equipment items.

We also note that page 2-2 of Attachment 2 discusses certain differences between the specification doses presented in Attachments' and 4 of this submittal and the specification doses listed on the System Component Evaluation Worksheets (SCEW sheets) presented in Attachments 4 and 5 to our letter No. AEP:NRC:0578B, dated June ll, 1982.

The reasons for these differences are also described in Attachment 2 to this submittal.

Since the specification doses for many electric equipment items were revised during the process of preparation of this letter, we have not yet had time to apply our quality assurance procedures to the new calculational results.

A quality assurance review of these results and an ongoing review of the radiation specification doses for the Donald C. Cook Nuclear Plant will be conducted following transmittal of this submittal.

If these reviews uncover the need for any additional changes, we will advise you of those changes by separate letter.

This document has been prepared following Corporate Procedures which incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.

Very truly yours, MPA/dam cc:

(attached)

M P. Al ch Vzce President

I Mr. Harold R. ~on

~

AEP:NRC:0578Z CC John E. Dolan W. G. Smith, Jr. - Bridgman R. C. Callen G. Charnoff E. R. Swanson, NRC Resident Inspector - Bridgman

ATTACHMENT 1 TO AEP:NRC:OS78I GENERAL DESCRIPTION OF ICE CONDENSER CONTAINMENT DONALD C.

COOK NUCLEAR PLANT UNIT NOS. l AND 2

1-2 General Description Of Ice Condenser Containment The ice condenser reactor containment system is divided into three major compartments the reactor coolant system or lower compartment, the upper compartment, and the ice condenser compartment.

Figures l-l through 1-3 present the general boundaries of these compartments.

These Figures also show the dead-ended compartments within containment whose air volumes are not displaced by steam into the upper compartment during a Loss-of-Coolant Accident (LOCA).

The lower compartment completely encloses the reactor coolant system equipment and associated auxiliary systems equipment.

The upper compartment contains the refueling canal, refueling equipment, and the polar crane which is used during refueling and maintenance operations.,

The upper and lower compartments are separated by a low leakage barrier (e.g., the operating deck) to minimum steam bypass between the compartments during a LOCA.

The dead-ended volumes are adjacent to the lower compartment and include the auxiliary pipe tunnel, the accumulator compartments, and the instrument room.

The ice condenser compartment, which contains the borated ice provided to quench the energy released during a LOCA< is in the form of a completely enclosed and refrigerated annular compartment which is located radially between the reactor coolant system compartment and the outer wall of the containment and, in elevation, generally above the operating deck.

That portion of the ice condenser which extends into the lower compartment has a series of hinged doors exposed to the atmosphere of the lower containment compartment.

For normal plant operation,'hese doors are designed to remain closed.

At the top of the ice condenser is another set of doors exposed to the atmosphere of the upper containment compartment.

These doors are also de'signed to remain closed during normal operation.

The ice bed is held within the ice condenser in baskets arranged to promote heat transfer from steam to ice should the condenser be needed to serve its function.

A refrigeration system maintains the ice in the solid state.

Suitable insulation

~

'urrounding the ice condenser compartment minimizes heat transfer to the ice condenser enclosure.

In the event of a Design Basis Accident (DBA) such as a

LOCA or a Main Steam Line Break (MSLB) within containment, the door panels located below the operating deck open due to the pressure rise in the lower compartment.

This allows air and steam to flow from the lower compartment into the ice condenser.

The resulting buildup of pressure in the ice condenser causes the door panels at the top of the condenser to open, allowing for air flow into the upper compartment.

The steam condenses quickly upon entering the ice condenser, thus limiting the peak pressure in the containment building.

The major factor which sets the peak pressure reached within containment as a result of a DBA is the compression of air displaced 'by steam as it flows from the lower and ice condenser compartments to the upper compartment.

Therefore, since the peak pressure is related to the ratio of volumes for the various containment compartments, judicious selection of compartment volumes during the containment design process can alter the predicted peak pressure until any desired value is achieved.

1-3 It is also noted that condensation of steam within the ice condenser causes a pressure differential between the upper and lower compartments which results in a continual flow of steam from the lower compartment to the condensing surface of the ice.

This helps reduce the time that the containment is at elevated pressure.

Performance Criteria For The Ice Condenser Containment The performance of the ice condenser containment is demonstrated by results and analysis of ice condenser tests performed on a full-scale section test at the Westinghouse Waltz Mill Site.

These tests confirmed the ability of the ice condenser to perform satisfactorily over a wide range of conditions, exceeding the range of conditions that might be experienced in an accident inside the containment building.

The ice condenser containment performance has been evaluated by testing the effect of certain important parameters.

A partial list of parameters tested include blowdown rate, blowdown energy, deck leakage, compression ratio, drain performance, ice condenser hydraulic diameter, dead-ended

volumes, and long term performance.

Analytic models have been developed to correlate. and supplement these test results in the evaluation of the containment design.

The results indicate that the analytic models are conservative and that the performance of the ice condenser containment is predictable relative to these variables.

The energy absorption capacity of the ice condenser is at least twice that required to absorb all of the energy that can be released during the "initial blowdown phase of the reactor coolant system for all reactor coolant pipe break sizes up to and including the hypothetical double-ended severance of the reactor coolant piping, or for any steam system pipe break size up to and including the hypothetical severance of the main steam line inside containment, without exceeding the containment design pressure.

Steam bypass of the ice condenser during a postulated reactor coolant system blowdown is to be avoided.

The operating deck and any other leakage paths between the lower and upper compartments are reasonably sealed to limit bypass steam flow to a low value previously approved during the design phase of the Donald C. Cook Nuclear Plant containment design process.

For the containment, the analysis considered bypass area as composed of two parts -- a conservatively assumed leakage area around the various hatches in the operating deck, and a known leakage area through the deck drainage holes for containment

spray, located at the bottom of the refueling canal.

Flow distribution to the ice condenser for any reactor coolant system pipe rupture that opens the ice condenser lower inlet doors, up to and including the hypothetical double-ended severance case, is limited such that the maximum energy input into any section of the ice condenser does not exceed its design capability.

The door port flow resistance and size provides for flow distribution for breaks that fully open the inlet doors.

For breaks that only partially open the inlet doors to the ice condenser, the lower inlet doors act to proportion flow into the ice bed to limit maldistribution effects.

Analysis of the ice condenser reactor containment performance has shown that the ice condenser alone is capable of preventing containment overpressure during the initial blowdown of the reactor coolant system or secondary side system within containment, such that containment spray is not a requirement for overpressure protection.

However, extremely small blowdown rates would not generate a differential pressure'ufficient to open the ice condenser lower inlet doors.

In this case, the energy release (even at an assumed small rate) would eventually require containment spray operation to prevent overpressure.

Another case has been examined where it is postulated that a Small Break LOCA (SBLOCA) precedes a larger break accident.

The larger break accident is assumed to occur before all of the coolant energy is released by the SBLOCA (i.e.,

a double accident).

During the SBLOCA

blowdown, some quantity of steam and air will bypass the ice condenser and enter the upper compartment via leakage through the operating deck.

The important design requirement for the case of the double accident is that the amount of steam leakage into the upper compartment be limited during the SBLOCA phase of the double accident so that only a small increase in final peak pressure results for the second part of the double accident.

In general, the resultant peak pressure will be a function of both break sizes, the time between breaks, and the steam bypass during both phases of'the double accident.

The containment spray system is used to limit the partial pressure of steam in the upper compartment due to deck bypass.

The key elements which determine the double accident performance of the ice condenser are the lower inlet

doors, which open at low differential pressure to admit steam to the ice condenser and limit bypass flow of steam to the upper compartment<

and the containment sprays which condense the bypass flow of steam and limit the partial pressure of steam in the upper compartment to a low value.

The containment spray set point actuation pressure has been set at, 3

psig to limit steam partial pressure to less than 2 psia in the upper "compartment for the double accident case.

After a LOCA, the ice condenser has sufficient remaining heat absorption capacity such that, together with the containment spray

system, subsequent assumed heat loads are absorbed without exceeding the containment"design pressure.

The subsequent, heat loads considered include reactor core and coolant system stored heat, residual heat, substantial margin for an undefined additional energy release, and consideration of steam generators as active heat sources.

The primary purpose of the containment spray system is to spray cool water into the containment atmosphere in the event of a LOCA>

thereby ensuring that containment pressure cannot exceed the containment design pressure.

Protection is afforded for all pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant pipe.

Adequate containment heat removal capability for the ice condenser containment is provided by two separate full capacity containment spray systems.

The containment spray system is designed based on the conservative assumption that the core residual heat. is continuously released to the containment as steam, eventually melting all ice in the ice condenser.

The heat removal capability of each spray system is sized to keep the containment pressure below design pressure after all the ice has melted and residual heat generated steam continues to enter the containment.

1-5 Additional information on the design and function of the Donald C.

Cook Nuclear Plant ice condenser containment can be found in the Final Safety Analysis Report (FSAR).

Xn particular, the FSAR provides a

copious amount of information on structural adequacy during seismic

events, containment integrity during postulated blowdowns of the primary and secondary coolant systems, details regarding how containment isolation is achieved during DBAs, etc.

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Figure 1-3 Sectional Elevation Plan "D-D" Ice Condenser Containment

ATTACHMENT 2 TO AEP:NRC:0578I DESCRIPTION OF SHIELDING CODE FOR CYLINDRICAL SOURCES DONALD C COOK NUCLEAR PLANT UNIT NOS.

1 AND 2

1

)

UI

2-2 This attachment contains a copy of a user's manual for a computer program entitled "SHLlGG".

SHLlGG was programmed,

debugged, benchmarked, and utilized in the electric equipment environmental qualification program for the Donald C. Cook Nuclear Plant Unit Nos.

1 and 2 by personnel of the American Electric Power Service Corporation.

The user 's manual describes in detail the basic models which SHLlGG uses in computing gamma and/or beta doses at an observer/detector located inside a cylindrical source, outside a cylindrical source, or in intervening shields (including an assumed pipe wall if a pipe containing recirculating sump fluid in the recirculation phase of a LOCA is the assumed cylindrical source).

In general, SHLlGG may be used for applications in which the dose rate at an observer is desired and the geometry of the source is other than a cylinder.

More specifically, for the case of equipment radiation qualification calculations, the subcompartments of the ice condenser containment were modeled as cylindrical sources within which equipment (i.e., observer/detector) was located.

Selections for subcompartment cyclindrical model dimensions (e.g.,

length or height and radius of an equivalent cylinder for each subcompartment) are described in more detail in Attachment 3 to this letter.

Although the user's manual correctly describes the SHL1GG program in use today, we have made minor modifications to the actual coding since the time the user's manual was issued.

In particular, a

calculational error was discovered in late 1982 with regard to the utilization of a Simpson's Rule 'approximation for the evaluation of the Sievert Integral (a factor in the dose rate equation).

This has been corrected and taken into account in computing the beta attenuation factors presented in Attachment 3 to this submittal.

The specified doses for equipment inside containment, as given in Attachment 3 to this submittal, are therefore slightly different than those presented in Attachments 4 and 5 to our letter No. AEP:NRC:0578B, dated June 11, 1982.

Furthermore, as noted in Attachment 4 to this letter, SHL1GG was not used in computing radiation doses for equipment, items outside containment.

Rather, a computer code entitled "NSLSHL3" was used in those calculations.

NSLSHL3 is, in effect, an earlier version of SHLlGG, and thus utilized much of the same methodology.

We note, however, that review of the earlier work on equipment outside containment has indicated that some of the NSLSHL3 output was misinterpreted in computing radiation doses (i.e.,

doses from a 14" outside diameter pipe were used for some equipment items, rather than the limiting doses from an 8.625" outside diameter pipe).

The specified doses for some equipment items outside containment, as given in Attachment 4 to this submittal, are therefore higher than those presented in Attachments 4 and 5 to our letter No. AEP:NRC:0578B, dated June ll, 1982.

The radiation qualification doses for electric equipment, important to safety (both inside and outside containment) are undergoing continual review for the Donald C.

Cook Nuclear Plant.

Revision No.

0 Date:

December 31, 1980 SHL1GG:

A Shielding Code for Cylindrical Sources Developed By:

v

~D /2 Bi/gQ G. Garner Date Verified By:

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P.

Manno Date Approved I. Castresana ate

(

Table of Contents I.

Abstract

~Pa e

II.

Introduction III.

Model A.

Geometry B.

Gamma Dose C.

Beta Dose D.

Source Strength E.

Decay of Isotopes and Integrated Dose IV.

Code Description and Options V.

Input Description A.

Input Requirements B.

Criteri'a for Choosing NRAD and NANGL 13 15 18 19 20 23 23 27 VI.

Required JCL VII..

Test Cases A.

Case 1

Dose Calculation - Time Zero B.

Case'1 Dose'Calculation

- Verifi'cation of Isotope Decay and Case Summations and Summaries 29 31 34 C.

Case 2 Dose'Calculati'on

- Time Zero VI'II.

D.

Case 3 Dose Calculation - Time Zero Benchmark. Case 49 53 I.X.

Re ferences 58 Appendix A.

Evaluation of F (1;b)

Appendix K.

Version SHL3 Appendix C.

Final Modifications of November 20, 1980 and December 16, 1980 60 62 63

C J

~ I

List of Fi ures 1.

Geometry - Observer Inside Pipe 2.

Geometry - Observer Inside Pipe Wall 3.

Geometry Observer Outside Pipe 4.

Calculation of c.

J

~Pa e

List of'Tables

,l.

Summary of Test Cases Run and Options Used 2.

Summary of Isotope Information 3.

Comparison of Scaled Code Results and Reference 8

Results -'aoea Dose

~Pa e

32 33 56 4.

Comparison of Scaled Code Results and Reference 8

Results - Beta Dose 57

SHLIGG:

A Shieldin Code for C lindrical Sources Abstract SHLlGG calculates dose rate and time integrated dose due to a cylin-drical pipe contai.'ni.ng a

gamma and/or beta source.

The source is distributed in water,

steam, a steam-water mixture, or air.

Gamma dose is calculated inside the pi.pe, outside the pipe, or in the pipe wall.

Beta dose is only calculated inside the pipe; beta dose beyond the pipe wall inner surface is expected to be small compared to the corresponding gamma dose.

For beta dose i,ns i de the pipe, SHLIGG decides whether the sour ce may be treated as an infinite cloud for each beta particle.

If'so, infinite cloud results are appli.ed.

I'f not, SHLIGG divides the source into a number of prismati.c elements.

Each element is treated as a line source, and a dose rate i.s calculated.

The indivi'dual dose rates are summed to produce a total dose rate for the entire source..

Beta and gamma doses are tabulated separately.

For gammas, buildup in the source, pipe wall, shield, and surrounding air is accounted for.

II.

Introduction SHL1GG is a FORTRAN IV computer program for calculating dose rate due to a cylindrical pipe containing a

gamma and/or beta source.

The source is distributed in water,

steam, a steam-water mixture, or air.

Gamma dose is calculated inside the pipe, outside the pipe, or in the pipe wall.

Beta dose is only calculated inside the pipe; beta dose beyond the pipe wall inner surface is expected to be small compared to corresponding gamma dose.

A For gamma dose outsi'de..the

pipe, a number of shields may be present I

between the source and observer.

The shields must be parallel to the cylinder axis and perpendicular to the perpendicular line joining the observer and this axis.

The shi,elds must'be large enough that the observer is fully shielded

(.ie.,

a line drawn from an arbitrary source point to the observer must pass through the shields);

In all cases, self-shielding of the source is considered.

Shielding due to the pipe 'wall is considered where appropriate.

Buildup of gamma flux in the source, and, i'f appropr iate, in the pipe wall, shields, and surrounding air is accounted for using a Taylor buildup factor (Reference 3, p. 415).

For beta dose inside the pipe, SHL1GG decides whether the source may be treated as an infinite cloud.'his decision is made for each individual beta source.

If so, a dose for that particular beta is immediately calculated using infinite cloud formulation.

If not, the source is divided into

~

prismatic elements.

Each element has sectors of two concentric cylinders and two radial planes forming its boundaries.

The division is done uniformly; the user specifies the number of radial divisions and the number of angular divisions.

Each prismatic element is treated as a line source.

Gamma dose rates are calculated for individual elements for each energy and summed to obtain total gamma dose rate.

Beta dose rate is calculated, where appropriate

(,ie.,

when dose inside the pipe is desired and the source may not be treated as an infinite cloud), in a similar manner for individual beta particles and summed (each beta parti'cle has a given energy probability density, maximum

energy, and average energy).

Formulation for beta attenuation analogous to that for gamma attenuation is used as described in Reference 1.

For both beta and gamma doses relaxation length is based on that portion of the per-

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pendicular distance from the particular prismatic element to the observer passing through each material.

E l

Dose rate is initially obtained at time zero.'Dose:rate at any later ti'me step is obtained by decaying the dose rate due to each isotope at the previous time step by the amount the respective isotope decays in the time e

interval.

I'n addition, time integrated dose is calculated at each time step.

A simple exponential decay i's assumed; dose due to product isotopes is neglected.

According to Reference 2, this effect may be accounted for by multiplying the above doses by 1.3.

Additional work has shown that this factor is valid for gamma dose, Gut not necessarily for beta dose (Reference 13).

III. Model A.

~Geometr The geometry for the observer/detector point inside the pipe, inside the pipe wall, and outside the pipe is shown in Figures 1 to 3'espectively.

For the case of the observer/detector outside the pipe, shields may be present and are perpendicular to line CO.

All remaining space outside the pipe is assumed to be filled with air.

The jt" prismatic element is shown, at location (.R.,o.).

For a' J

single beta or gamma energy Ek assume this prismatic element has a line intensity SL

.k (in Ci/cm).

From Reference 3, p.

348, the dose rate D.k due to this element and'energy is:

(~)

where K is a conversion factor equal to 3.7 x 10 dis/Ci-sec and'0

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= buildup factor for'energy Ek and dimensionless distance b..

J k

jk'OF)k F(O,b)

= dose conversion factor at energy Ek in Rem-cm -sec/hr for gammas, and Rad-cm sec/hr for betas.

2

-bse~

e'he dimensionless distance b

is given by gk bjk (2a) observer inside pipe

~jk

~~~k cj + ~p~k Jj (2b) observer inside pipe wall A, y. Cg + W~,g g. y ~

atr> K, atr N

+ g My,+5&c $.

(2c}

observer outside pipe '(with

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~~>> >= linear attenuation coefficients of source,

pipe, and surrounding air respectively, at energy Ek.

= distance traveled through surrounding air for; observer outside pipe

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Remai.ning quanti.ti;es are defi,ned in Figures 1-3.

~f Ro Ri Lo RpBS LpBS R-, and 0. are given, then a, c., d.,

J J

J Jl.,

Q2. and/.

may be calculated.

In all cases, a., Ql., and Q>.

are given by:

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t/<

Ps

( Rg

+ R~8~ -2 g g~Q~ gcsg 8.)

(3) 9 +

gj

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J (4) 9I s cail J

Note that LOBS 0 0 measured along the cylinder and L

4 0 OBS measured antiparallel to the cylinder (see Figures 1-3).

Then e.,

e

- y 0 lj'j measured clockwise, and 91'.,

02 g 0 measured counter clockwise as required J

J (see Figures 1-31.

c is obtained by reference to Figure 4.

The'oordinates of E are (R

d,oglsl g

Qj>> Ig.).

The. coordinates of H are given by the solution to x ay~= R (6a) y Rd Sin 8 R cps & R

(

oSs)

J J

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R'~(t+ w~) m~ P~g~

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In (7a), the positive sign is chosen, as the larger yalue. of x is required (ie., the value corresponding to H, and not H', in Figure 4).

In (7b), the positive sign is chosen for 0 4 9 g +, and the negative sign is J

chosen for )l 4 9 + 2'll'.

Then c

is given by c> = [ (y.< g.> cos 8) + (

> g> s>n e>)J (8).

To calculate d, the quantity c

+ d. is obtained by replacing R.

J J

1 by R

in Figure 4 and equations (6) - (8).

Knowing c and c

+ d, d.

J J

J' is obtained.

To otitain g, it is seen from Figure 4

(ga)

S/n 4 = S]n 8.

gg J

J Yz co> p = P/

(

. Gin 9>)

(gb)

(gc)

SeC.

$ = Ll ( ~ Stn8) 3 B.

Gamma Dose Gamma dose is calculated for each gamma energy in accordance with equation (1)..

Linear attenuation coefficients for various materials as a

function of energy are given i'n Reference 4, p.

82.

e

Oui.ldup factors for energy E< are (loa) outside pipe (lob) inside pipe wall (loc) inside pipe where:

buildup factor in source Sqj 5 A

e 8'9 K

(>>a) observer outside pipe or inside pipe wall k

CPg J

/

h kg ~Spk J

(>>b) observer inside pipe J

(

g )~%~~A<

3

(>>c) observer inside pipe wall or outside pipe

g..

bui1dup factor in shield i n

+tc~itc6~ sec qb I

~

I N:Ag~+sec 4 observer outside pi pe, and shi el ds are present.

Ak,o<k, and gk, for various materials as functions of energy, are given in Reference 3,

pp. 416-423.

S is approximated as 1.05 (this may be verified to be conservative for most problems of interest using Reference 4, p. 527).

For steam, B

is also taken to be 1.05.

For steam-water

mixtures, B is calculated for steam and for water and is weighted by quality.

Oose conversion factor as a function of energy is given in Reference 3,

p.l9 (final dose is in R/hr).

Curve fits to this figure for various energy ranges are used in SHL1GG.

C.

Beta Oose Beta dose is calculated inside the pipe only; beta dose beyond the pipe wall inner surface is expected to be small compared to corresponding gas+a dose.

Seta dose calculation may be formulated analogously to gamma dose cal-culation (Reference 1, pp. 625-629).

This is due to the particular analytical forms of beta spectra and electron scattering and absorption cross sections.

Beta dose is calculated as in equation (1).

Linear attenuation co-efficient is {Reference 1, p. 628)

/7 (7

~

I Ip where E=-

is the maximum possible energy for the given beta particle, and P is the material density.

Average beta energy (for a given beta particle) is approximated as one-third maximum beta energy (Reference 1, p. 540).

Then g IlV In this formualtion, there is no Guildup of beta flux.

Buildup factor in equation (I } is i'gnored.

Oose conversion factor is given by (14) where W for beta attenuation is evaluated using (13) with average energy Ek.

This is conservative, as this ~ is for total attenuation (absorption pIus scattering}; actually only $ for absorption should be used. ~ for absorption is not'vailable.

To obtain beta dose in Rads/hr (DF )= ~ Effimcsss

[fd

~ stfm m < j 3

] Rod 3@o<

s<c pm J Cms Sco

= &.7d'7 }clo E

(15)

I

. ~

I'J

m m

'For a given beta average energy, if infinite cloud assumptions

apply, dose is given by (Reference 5)

(16) instead of equation (1).

In equation (16),

Dk

. = air dose rate in Rads/sec Ek

= beta particle average energy in HeV S

1

= activity concentration in Ci/m For an arbitrary infinite cloud source medium where:

P ;r

= densi,ty of air

= 1.293 x

10 g/cm at standard conditions (Reference 3, p.

19)

= densi'ty of source in g/cm D.

= dose in source in Rads/sec.

Inserting P.and altering the units so that 0

is in Rads/hr and S

k v,l in Ci/cm produces (1S)

The source may be treated as an infini'te cloud for a particular beta energy p

es of that energy produced at the observer point are effectively if all beta arti'cles absorbed in the source.

Relative to the observer, the source looks like an i.'nfi.nit.e medium.

The infinite cloud model as used here has meaning only for the observer poi.nt inside the source.

The criterion for treating the source as an infinite cloud is that a

beam of beta particles of the'respective average energy emitted at the observer point must be reduced'to 1% of its initial intensity over the shortest distance between the'bserver and the pipe walI.

In this case con-tributions to the beta dose'at the observer point due to source points beyond this di:stance are neg1 igi51e.

The shortest di.stance between the observer and the pipe wall is Then the above criterion may be stated PXp p Af (

P.1 s) 3 ( Q.Q)

(20) where +> = linear attenuation coefficient for betas as given in equation (13/.

This gives RTS >

g. Co (21)

When (21)

(.

) is satisfied, the source may be treated as an infinite cloud for betas of the particular energy, and (18) may be used.

~Sh The line intensity in equation (I) is obtained from (22) where:

S 1

= activi ty density of isotope 1 (in Cigcm3) lk

= fracti"on of activity of isotope 1 comprising garnna energy or average beta energy Ek A>

cross-sectional area of element j

R h,RBO j

Activity dens ity i s obtained from (23) where:

St 1

= total activity of isotope 1 (in Ci)

= fraction of this total activity released

= volume into which this activity is released.

Equation (1) is summed over elements (j) and energies (k) to get total dose rate.

Note that in the output, those values labeled activities (both activities for each isotope and total activity for all isotopes) do not incorporate the fractions fl.

Those values labeled activity concentrations do incorporate the fractions f.

1'.

Deca of Isoto es and Inte rated Dose Simple exponential decay is assumed; dose due to product isotope is neglected.

According to Reference 2, this effect may be accounted for by multiplying the obtained doses by 1.3.

It is shown in Reference 13 that this is valid for gamma

dose, but not necessarily for beta dose.

For each isotope, a decay constant must be input.

Activity, activity concentration, gamma dose for a given energy, and beta'dose for given energy at time t + ht are obtained from corresponding quantities at time t via (24) where c i,s the decay-constant, and X is any of the above four quantities.

Integrated beta or gamma dose for a given energy at time t +At is obtained from where Y is i.ntegrated dose and X is dose rate for the respective gamma or beta energy.

IV.

Code Descri tion and 0 tions SHLlGG consists of a main program and the ll subroutines

P01, PDZ,

PD3, DSR, F, GXP,'BLDPF,

ABSRP, SNTP, BLOUP, and ORY (in addition, there is a block data subroutine that initializes common block 83; this block is common to HAIN, P01, POZ and PD3).

The main program reads and prints input, calcu-lates decay of isotopes and dose rates over time, calculates integrated

dose, and summarizes individual and total doses at each time by energy or isotope as desi.red by the user.

PD1 calculates gamma dose rate outside the pipe or inside the pipe wall at time zero for each gamma energy.

P02 calculates gamma dose rate in-side the pipe at ti'me zero for each gamma energy.

P03 calculates beta dose rate inside the pipe at time zero for each beta average energy.

OSR(EN)

I i'

- 21 evaluated gamma dose conversi.on factor for engy EN, For this calcula-tion, curve fits to various portions of Figure 2,l,of'Reference 3 are used.

F(THETA,B). evaluates.

the Si'evert integral function F(o,b).

BLDPF(MATH,EO,Y, YL,YT) and BLDUP(MATH,B,Y,YL,YT,X)calculate buildup factor for material" MATH, di'mensionless length 8, energy EO and quality X

(in the case of steam-water mixtures).

Y, YL, and YT are the quantities Ak, o( k, and fk respectively i'n equation (11); these depend only on energy (and not on 8).

The calculation of buildup factor is divided into a por-tion which depends only on energy (BLDPF) and a portion which r

depends on distance traveled and energy (BLDUP).

Execution time is reduced by calling BLDPF only once for each gamma energy.

This portion of the calculation need not be done for each source element.

BLDUP must be called once for each gamma energy and source element.

ABSRP(MATH,EO) evaluates gamma +

for material MATH and energy EO.

DRV(AIS,DTT,NI'SOT,FCTD) evaluates fractions by which istotpes decay over the given time step (ie., evaluates p

~

1, 2,...,

NISOT),

where DTT = time step si'ze AIS = array containing isotope informati'on AI'S(S,J),

= ~

= decay constant for isotope J

NISOT = number of isotopes FCTD(J} = e'

(,J

= 1,HISOT)

SNTP(X,ARG,VAL,Y,NDIM) does linear interpolation.

ARG(NDIM) and VAL(NDIM) are arrays of the independent and dependent variable respectively.

X is the argument for which the output Y is desired.

Y is obtained by linear inter-pol ati on.

GXP eva 1uates the function 8.-(~tco~ c3 b

If aas e.

is large enough to cause an underflow, GXP is set to zero.

g t h

4 0

The source,

pipe, and shield materials are limni.ted to water, air, iron (represents all steels}; concrete, and lead.

Water includes steam and steam-water mixtures; quality must be input (steam, x

= 1; water, x = 0; steam-water mi.xture, 0< x< 1J.

When dose outside the pipe is calculated, the pipe is assumed to be surrounded by air (except at shield locations).

This air need not have the same density as a possible air source material.

At 4.'ach time step, the code will print out dose rate and cumulative dose for each individual beta average energy and gamma energy, totals for each isotope, or totals for all isotopes.

A different option may be in effect in each print interval.

A print interval is defined as a period of time with constant time step size and print option.

There may be many print intervals, each with its own time step size and print option.

At each time step, total activity and activity concentration, total dose rate and integrated

dose, and ratio of activity concentration to dose rate are always printed.

Beta and gamma doses are always printed spearately, as gamma doses are given in R/HR and beta doses are given in RADS/HR.

When dose is calculated inside the pipe, the user may specify beta dose is to be evaluated.

Beta dose is calculated inside the pipe only.

The code may be run under two options.

With one option a single. case is run, while with the other option a number of cases are run.

A case is defined as a single problem with a specified set of physical and geometrical parameters.

For the former option, any number of shields may be present.

For the latter option, a single shield is present, and the shield thickness is incremented by a given amount for each case.

The initial thickness may be zero (if no shield is present in the first case).

In all cases, the user must insure that the shield is thin enough to fit between the pipe and the observer (if not, an error message is printed out and execution stops).

H All calculations are done in double precision.

23-

~V.

,In ut Descri tion A.

In ut Re uirements

')

Card 1

(2 cards)'(lOAB}

These cards contain information identifying the source reference of beta and gamma energies.and decay fractions.

This informati'on is printed.

2}

Card 2 (F10.5)

SYSVOM SYSVOH = system volume in cm3 3)

Card 3 (2A8),

SCHAT, PIPMAT SCHAT = source material name PIPHAT

= pipe material name Names are limited to:

MATER (includes steam-water mixture),

IRON, CONCRETE, AIR, and must be left justified.

Any other input will produce an error

message, and execution will stop.

4)

Card 4 (3F10.5)

RO, RI',

LO RO

= pipe outer radius (cm)

RI = pipe inner radius (cm}

LO = pipe length (cm) 5},

Card 5 (2IS)

NISOT, BTOPT NISOT = number of isotopes comprising source (maximum of 100)

BTOPT = beta option parameter 1

calculate beta dose 0

don't calculate beta dose

. 6)

Card 5

(NI.SOT Cards)

Each. Card:

j}

I'sotope name, activity (Ci)

(total activity for this isotope),

number of different gamma energies emitted, fraction of total activity of this isotope released'nto volume SYSVON, decay constant (sec

},

number of different average beta energies emitted - A8, F12.3, I5, 2F12.3, I'5.

I'sotope name is a character string identifying the isotope; it should be left justified.

ii}

Gamma energy (MeV), fraction of gammas emitted with this I

energy, gamma energy, fraction of gammas emitted with this energy,... SF10.3.

There are as many entries as different gamma energies for this isotope;.information should be continued onto as many cards as required.

The total number of gamma energies for all isotopes must be less than or equal to 1000.

iii} Beta average energy (MeV), fraction of betas emitted with this average

energy, beta average energy, fraction... SF10.3.

Theie are as many entries as different beta energies for this isotope; information should be continued onto as many cards as required.

The total number of beta energies for all isotopes must be less, than or equal to 500.

?}

Card

? (.2IS}

NSHLD, SHLDOP NSHLD = number of shields (20 maximum; may be zero)

~

~

la

~ ~ SHLDOP

= shi:eld opti.on SHLDOP

= 0 1 case possibly many shields SHLDOP

=

1 possibly many cases,'I.shield.

For this option,'SHLD must be l.

8)

Card 8 -

NSHLD cards (A8,2F10.3}

Included only if NSHLD>0 Shield ma teri a 1 name, shield thickness (cm), shi el d dens ity

('g/cm3)

Shield materi'al name must be chosen from the names given in Card Set 1 descr iption, and must be left justified.

If SHLDOP

= 1, shield thi'ckness is the initial thickness (case 1).

9}

Card 9 - (I5, F10.3}

Included only if SHLDOP

=.

1

NOPT, SING NOPT

= number of cases SINC

= shield thickness increment (cm) 10)

Card 10 (.4F10.3)

SCDEN, X, PIPDEN, AIRDEN SCDEN

= source material density (g/cm3)

X = source quality (ignored unless source is water; o<~~ < ~

PI'PDEN

= pipe material density (g/cm

)

AIRDEN = density of air surrounding pipe (g/cm

)

Note that quality is specified only for the source.

If water is used as a shield, it must be liquid.

1 1 )

Card 11 (2I5)

NRAD, NANGL NRAD = number of radial divisions

NANGL = number;of angular divisions 12}

Card 12 - (2F10.3)

ROBS, LOBS ROBS

= perpendicular distance from source centerline to observer (cm)

LOBS = distance along source from one end of source to observer (cm).

LOBS may be measured from either end of the source, as the entire system is symmetric with respect to the source midplane.

LOBS> 0 indicates LOBS is measured parallel to the source, while LOBS< 0 indicates LOBS i's measured antiparrallel to the source (see figures 1-3).

13}

Card 13 - NPRNT+1 cards i)

NPRNT E5 NPRNT = number of print intervals (20 maximum}

if) NPRNT cards.,

one for each print interval - giving en<<i<<

for each print interval (sec),

number of steps in each print interval, print option F10.3, 2E5 0

print, doses for each energy Print Opti'on

=

1 print only totals 2

print totals for each isotope

8.

Criteria for Choosin NRAD and NANGL NRAD and NANGL must be chosen such that, for the lowest energy gamma and beta, there is not appreciable attenuation over the extent of any element.

This is because each element is treated as a line source concentrated at the element center and shielded by the element source I

material.

If there is appreciable attenuation, lumping will cause radi-ation from portions of the element closest to the observer to be attenuated more than it should be.

Resulting doses calculated will be non-conservative.

The largest elements will be at the periphery.

They will be of extent (26a)

(26b) where:

b,s

= arc length subtended by element mid-section and other quantities are defined in Figures 1-3.

For attenuation to be small over the extent of an element (27a) chS

<~ I (27b) where: Q is the largest attenuation coefficient, corresponding to the smallest gamma and/or beta energy.

Substituting (26) into (27) gives

hid.AP>>

b (28a)

(28b) where b

= R.g(

max

, i max When beta doses are calculated, g for betas will generally max be the controlling factor, because beta attenuation is much greater than gamma attenuation.

However, in many cases beta attenuation will be so large that infinite cloud results will apply; then gamma attentuation will be the controlling factor.

The user should choose NRAD and NANGL based on (28) and the particular beta and gamma energies input.

It is sufficient that the inequalities be satisfied by a factor of 10.

It is also necessary that the element size be small compared to the distance from the element to the observer.

This requirement is significant when an observer is close to a large source.

As an example, assume a water source and 0. 1 t<eV smallest energy gamma.

This is the lowest gamma energy the code will distinguish; is taken as a constant equal to ~ at 0. 1 MeV (=0.167 cm

) for lower engeries.

Y'his is conservative.

Assume beta dose is not calculated.

Then

= O. /67 cm" b = o,lo7 R.

where R. is in cm.

Therefore 1

NRAD m) 0.167 R.

1 NANGLw> 1.05 R.

1

~

I ~

6-While satisfying (28) is sufficient for calculated doses to be conservative, it is not necessary.

In particular, for the observer at the center of the pipe, NANGL can be 1 with no loss of accuracy, due to symmetry.

For, the observer slightly off center, (28b) need not be satisfied.

For doses outside the pipe or in the pipe wall, (28a) and (28b) should both be satisfied.

C.

T ical Densities are Typical densities for the various source,

pipe, and shield materials Water (x=o)

Air Iron Lead 1.0 g/cm 3

1. 293 xl0 g/cm 7.83 g/cm 11.3 g/cm dJL SHLlGG source is stored in member SHLlGG of library SRCENSL.

To run SHLlGG, the following JCL may be used:

JOB CARD

//STEPl EXEC SRCE,LIB=NSL

//SRCE.DATA DD *

=H NSLKCHR VARIOUS =R,

=D,=A CARDS AND NEW DATA

//STEP2 EXEC SSFLG,C=C,SRC=NSL

//Gg.SRCE DD *

=M SHL1GG

//FT05F001 DD DSN=*. STEPl.SRCE. SYSUT2,DISP=(gLD,DELETE)

Note:

5 = letter 0

0

= zero The member NSLKOHR contains SHL1GG input.

A listing of this member is attached as Run ll.

The isotopes and activities are taken from Reference 10, Table 5-29-,

and Reference ll, Appendix A, Table A.2-1.

For Reference 10 activities, a

U mass of 8.86xl0 Kg is used (see Section VIII and Reference 11, Chapter 3 (Unit 1), Table 3.2.3-1).

In cases where the two references conflict, Reference 10 activities are used.

Decay con-

stants, gamma and beta energies, and fractions of total isotope activity

.comprising each energy are given in Reference 12 (this is indicated in the first two input lines).

It is assumed that lOOX of,the noble gases, 50K of the iodines, and lX of the particulates are released into volume SYSVOM.

These items constitute the bulk of the input.

With the exception of the fractional releases, these items generally do not change from run to run.

Remaining input items (lines 100-600 and 37200-37700 of NSLKCHR) must be input by the user for the desired case(s).

This may be accomplished by modifying NSLKCHR as necessary via

=Ri

. =A.

and

=o cards.

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SUBJECT

'III.'enchmark Case The beta and gamma dose rates inside a typical dry containment during the thirty day period following a LOCA are calculated by the code and compared with independently obtained results summarized in Reference 8, Appendix D (additional information is given in Reference 7, Appendix B.)

A reactor power of 3391 NWT is assumed.

The containment volume is 2.52 x 10 ft

= 7. 136 x 10 cm (Reference 7, p. B-2).

The containment 6

3 10 3

is modeled as an equivalent cylinder with height equal to inner diameter (Reference 8, p. D-7).

This gives an inner radius of 2248 cm and a.height of 4496 cm.

All remaining dimensions are taken from Figure A-2 of Reference 9.

The containment is assumed to be filled with air of density 1.293 x 10 g/cm The analysis in References 7 and 8 assumes the only fission products present are noble gases and iodines.

Both references assume that lOOX of the noble gases and 25K of the iodines are released into the con-tainment.

Total activities are given in Reference 10, Table 5-29, and Reference ll, Appendix A, Table A.2-1.

In cases where the two references conflict, Reference 10 activities are used.

For Reference 10 activities, a

U mass of 8.86 x 10 Kg is assumed.

This roughly corresponds to a

U02 mass of 216,600 ibm (Reference ll, Chapter 3 (Unit 1), Table 3.2.3-1).

Decay constants are given in Reference 12.

Gamma and beta energies, and fractions of total isotope activity comprising each energy are given in Reference 12.

The source is divided into 5 radial division and 18 angular divisions.

This may be shown to satisfy the criteria of section V. B.

For an air source, the maximum attenuation coefficient used by the code is (P/()

= 0.151 cm /g (see SUBROUTINE ABSRP listing).

Then 2

cm 2

-3 bmax (0. 1 51

)

( 1.293 x 1 0 ~) (2248 cm)

O 438g cm 2 Tfb

= 2. 758 Therefore, (28a) and (28b) are satisfied.

IC Dose

. is ca 1 cul a ted at one hour intervals for, one day, and then at one day intervals up to thirty days.

Print option 2

(total s for each isotope) is used

~

Al 1 input is summari zed in run 8.

Reference 8 accounts for iodine removal by plate-out.

In addition, Reference 8 may not use the same initial source strengths as are input to the code.

Therefore, by the code are scaled by I,REF. 8 I,CODE iodine and noble gas doses calculated and NG, REF. 8 HG, CODE respectively,

where, A

iodi ne activity used in Reference 8

t A<G R F 8 noble gas activi ty used in Reference 8

AI CODE 1 odi ne acti vity used by code AgG CODE nobl e gas acti vity used by code.

All activities are evaluated at the time step in question, and must be expressed in units such that the above ratios are dimensionless.

A> <0DE must account for 25K of total iodine activity.

Scaled code results and Reference 8 results are compared in Tables 3 and 4.

With the exception of the beta dose rate at t= 0 and t,=720 hrs (30 days),

scaled code results are higher than Reference 8 re-sults.

Heta doses agree reasonably well; discrepancies are most likely due to discrepancies in input beta energies and decay fractions.

Gamma doses agree:weU."up to.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Discrepancies are.most likely due to discrepancies in input gamma energies and decay fractions, and conserva-tive approximations in the code calculation of gamma dose.

TABL COMPARISON OF SCALED CODE RESULTS AND REFERENCE 8 RESULTS GAMES, DOSE VO ~+

COMVARGL)

Time Iodine Concentration (108Ci)

2. 06 Noble Gas Concen)ration 1 (10 Ci) 5.31 Iodine Dose Rate2 (105R HR) 8.95 Noble Gas Dose Rate>

(10 R HR) 8.40 Scaled Iodine Dose Rate3 (10 R/HR) 19.9 Scaled Noble Gas Dose Rate 3 (105R/HR) 20.72 Total Scaled Dose Rate4 (10 R/HR)

40. 62 Reference 8

Dose Rate5 (105 R/HR) 37.9 1.61 4.41 6.05 5.95 0.63 8.65 9.28 8.1 7ZO 0.83

0. 477

0. 193 0.0184 2.70

2. 023 1 ~ 239 0.0456 1.86 1.51

0. 77 0.54 0.221 0.296 0.01971 0.00887

0. 13 0.048 0.013 0.00114 2.02 0.622 0.317 0.01169

2. 15

0. 670

0. 330 0.01283

1. 82

0. 399 0.100 0.00395 Activities from Run 8.

Iodine activities multiplieg by 0.25.

2Dose rates from Run 8.

Dose rates scaled using Reference 8 activities from Table D-2, Columns 2 and 6.

Sum of Columns 6 and 7.

Dose rates from Reference 8, Table D'-5, Column 2.

Dose rates are reduced by a factor of 1.3 for comparison with code results (see Reference 8, p. D-7).

TABL COMPARISON OF SCALED CODE RESULTS AND REFERENCE 8 RESULTS BETA DOSE Time (HR)

Iodine Concentration 1 (10 Ci)

Noble Gas Concentration (108ci)

Iodine Dose Rate (105R/HR)

Noble Gas Dose Rate (10 R/HR)

Scaled Iodine Dose Rate (10 R/HR)

Scaled

'Noble Gas Dose Rate (10 R/HR)

Total Scaled Dose Rate4 (10 R/HR)

Reference 8

Dose Rate (10 R/HR) 96 720 2.06 1.61 0.83

0. 477 0.193 0.0184 5.31 4.41 2.70

2. 023 1.293 0.0456 20.79 15.08 6.42 3.28

• 0.909 0.07678 43.

09'0.54 8.60 4.68

2. 714

0. 08565 46.23 1.57 0.45 0.204 0.053 0.0044 106.3 44.40 11.50 5.39 2.907 0.1117 152.53 46.0 11.95 5.59 2.96 0.11610 182.5 25.9 ll.8 5.44 2.40 I

0. 146 I

1Activities from Run 8.

Iodine activities multiplied by 0.25.

-2Dose rates from Run 8.

3Dose rates scaled using Reference 8 activities from Table D-2, Columns 2 and 6.

Sum of Columns 6 and 7.

5Dose rates from Reference 8, Table D-6, Column 2.

Dose rates are reduced by a factor of 1.3 for comparison with code results (see Reference 8, p. D-7).

A

)

~ References 1.

Robley D. Evans, "The Atomic Nucleus",

McGraw Hill, New York, 1955.

2.

M. J. Kolar and N.

C. Olsen, "Calculation of Accident Doses to Equipment Inside Containment of Power Reactors",

Trans.

ANS, 22, 1975, pp. 808-809.

3.

Theodore Rockwell, Reactor Shieldin

Manual, Van Nostrand, Princeton, N.

J ~, 1956.

4.

John R.

Lamarsh, Introduction to Nuclear En ineerin, Addison-Wesley,

Reading, MA. 1975.

5.

"Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors",

Regulatory Guide 1.4, U.

S. Atomic Energy Commission,

June, 1974.

, 6.

Jan J.

Tuma, En ineerin Mathematics

Handbook, McGraw Hill, New York 1970.

7.

"Environmental gualifcation of Class IE Equipment",

IE Bulletin 79-01B, U. S. Nuclear Regulatory Commission, January 14, 1980.

8.

A. J. Szukiewicz, "Interim Staff Position on Environmental qualification of Safety-Related Electrical Equipment",

NUREG-0588, U.

S. Nuclear Regulatory Commission,

December, 1979.

I

c 59-

9. 'awrence J. Metcalfe, Walter J. Mings, John E. Hartman, and Allan C.

Crail, "CONTEMPT4/M002, A Multipcompartment Containment System Analysis Program",

TREE-NUREG-1202, EGRG:.

Idaho, Inc., February, 1978.

10.

"Radiation Analysis Manual, Standard Plant Model 412", Revision 3, Westinghouse Proprietary Class 2, Westinghouse Electric Corporation, Pittsburgh, PA, November, 1978.

ll.

Final Safety Analysis Report,'onald C,

Cook Nuclear Plant.

12.

D.

C. Kocher, "RADIOACTIVE DECAY DATA TABLES:

A Handbook of Decay Data for Application to Radiation Dosimetry and Radiological Assessments",

l U.

S.

Department of Energy (in press; tape with data available from D.

C.

Kocher ).

13.

. V.

P.

Manno, "Calculation of the Effect of not Including Decay Chains in Assessing Doses,"

NSL calculation RD 80-03, October 31, 1980.

It I!

I' A

endix A

Evaluation of F O,b The Sievert Integral Function, F(O,b), is defined by 8

bsec

&I P

(Al)

In all applications here, b~o and -W 4 & 6 II'his function is evaluated by the function subprogram F(THETA,B).

The following approximations are used:

b<l~

0( 9< ceo F(&ik)=

8 e (AZ)

This approximation is given in Reference 3, p. 408 for e<5 However, examination of curves for F(O,b)

(Reference 3, pp. 385-390) shows that for b <l, this approximation is good for e up to 60 (A3)

This approximation is conservative for all (O,b) in this range.

F(O,b) is overestimated by as much as"50Ã for O = 60 (Reference 3,

pp. 385-390).

3) b>i Oc gg ~

F(e,k) = ae'A4)

"a C

V

~

-C

~

~

~ r (Reference 3, p. 408)

4) I>>~

W <&<aO For this range of (G,b) a Simpson's Rule approximation is used with a step size equal to 0.0le (Reference 6, p. 212)

5) Q>~

4o i 8<go (A5)

(Reference 4, p. 439)

For 8<0, the result F(-o,b)

= F(o,b) is used (Reference 4, p.

439).

t g

~

1

gf ~l

~

7

~

I ~Appendix B

Version SHL3 An earlier version of the code, SHL3, has been used in several applications.

This version differs from SHL1 in that beta dose is calculated based on an infinite cloud source regardless of the cylinder size or beta energy.

For small cylinders or high energies this will be conservative.

ln addition, beta activity, energy, and dose input and output formats for SHL1 and SHL3 differ.

Run 9 reproduces case 1, run 1 using SHL3.

Comparison of runs 1.and 9 shows that SHL1 and SHL3 results for gamma dose are identical.

For beta dose (dose in water, run 9), run

9. is 0.3% larger.

This due to a constant in SUBROUTXNE PD3 of SHL3 being 0.3% larger than the corresponding constant in SHL1.

This error is insignificant, and doses inside a cylindrical source calculated by SHL3 are correct.

The purpose o'f this appendix is to verify SHL3 for previous applications.

SHL3 is not presently used.

A endix C

Final Modifications of November 20, 1980 and 'December 16 1980 Several modifications were made to the main program.

These changes relate to branching around certain DO loops.

An additional CONTINUE statement was added immediately following each respective DO loop.

The branch around the loop was changed to a branch to this statement.

Previously, the branch was to the DO loop terminal state-ment.

This can have adverse effects in IBM FORTRAN.

All changes are labeled ll/20/80 in the updated source listing (run 10).

Previous numerical results are unaffected.

Effective December 16,

1980, the source member name is changed from NSLSHLl to SHLlGG.

This is accomplished by the attached run 12.

ATTACHMENT 3 TO AEP:NRC:0578I DESCRIPTION OF CALCULATIONALPROCEDURE USED IN DETERMINING RADIATION DOSES TO ELECTRZC UIPMENT INSIDE CONTAINMENT

3-2 Introduction In order to calculate the total integrated dose to selected electric equipment inside containment, the upper volume, lower volume, fan/accumulator rooms, pipe tunnel, instrument

room, and total free volume were modeled into equivalent cylinders, based on volume information obtained from the Donald C. Cook Nuclear Plant Final Safety Analysis Report (FSAR), Appendix p.

The computer program SHL1GG,'escribed in Attachment 2 to this submittal, was then used with an appropriate source term to obtain the integrated doses and beta attenuation factors of interest.

These values were then applied to equipment specific calculations, based upon required operating times and the actual shielding installed at the plant.

Volume Nodalization For the overall containment, the total free volume's equivalent cylinder was determined to have a radius of 50.5 ft and a height of 145 ft (case 1).

Calculational results for this case are assumed to bound those for the upper volume, which was determined to be equivalent to a cylinder 50.5 ft in radius with a height of 84.4 ft.

Three models were developed to describe the lower volume.

First/

the annulus around the reactor cavity was treated as an equivalent cylinder with a radius of 17 ft and a height of 162 ft, into which airborne radiation is released (case 2).

A second approach which was deemed more likely to be conservative was to assume that the reactor cavity volume is a part of the lower volume.

This equivalent cylinder had a radius of 41.5 ft and a height of 40 ft, into which airborne radiation was released (case 3).

The third approach involved taking that portion of the lower volume which is expected to be submerged, and modeling it into an equivalent cylinder with a radius of 41.5 ft and a height of 18 ft (case 4).

With regard to the fan/accumulator rooms, they were modeled into equivalent cylinders of radius 10.5 ft and height 79 ft (case

5).

The results obtained for these cylinders were assumed to bound those for the instrument room

, which was determined to be equivalent to a cylinder of 10.5 ft in.radius and 67 ft in height.

The pipe tunnel was described as an equivalent cylinder of 7.5 ft in radius and 141 ft, in height.

Calculational results were obtained for the pipe tunnel both as if the tunnel had been flooded (case 6) and as if it had not been flooded (case 7).

For the earliest stages of a DBA>

a source term in air or an air-steam mixture,may be a more realistic assumption than that of instantaneous flooding.

~

g

~

3-3 Source Terms And Calculational Results Activities released into the containment for the assumed DBA were'valuated using values specified in Table 3-1.

This Table lists fifty-four (54) radioisotopes which have been identified as being characterisitic of the Donald C.

Cook Nuclear Plant core.

This list was compiled from two sources -- the Donald C.

Cook Nuclear Plant

FSAR, Appendix 14A, and the Westinghouse Standard Information Package, "Radiation Analysis Manual:

Standard Plant Model 412," Revisi.on 3, dated November 1978.

Using Table 3-1, it was a simple task to calculate the activities released into containment for the assumed DBA.

Using NUREG-0588 as a

basis of reference, 100% of the core inventory of noble gases, 50% of the core inventory of radioiodines, and 1% of the core inventory of particulates was assumed to be released into the containment free volume after flood-up (e.g.,

1,157,408 ft ).,

These activities are presented in Table 3-2.

For submerged components, 50% of the core inventory of radioiodines and 1% of the core inventory of particulates was assumed to be distributed into approximately 715,500 gallons of water (assuming a

maximum flood-up elevation of 614 ').

Halogen plate-out on containment surfaces and.halogen removal by containment sprays and borated ice were not assumed for conservatism.

In order to examine radiation attenuation in concrete, SHL1GG was used to calculate the integrated gamma dose at the center, the inner

edge, and the outer edge of a cylinder with an inner radius of 50.5 ft, an outer radius of 52 ft, and a height of 145 ft.

These calculations indicated that an 18" thickness of concrete (density

= 2.35 grams per cubic centimeter),

which is a typical concrete thickness between compartments, would attenuate all of the beta particles and a substantial portion of the gamma rays.

More specifically, integrated gamma doses were compared at various times between the three locations of'concern.

For the accumulated one-year

dose, the value at the inner edge of the cylinder was approximately 52.0% that at the centerline of the cylinder, and the value at the outer edge of the cylinder was a mere 0.051% that at the centerline of the cylinder.

Therefore, it was concluded that, the dose contribution from one compartment would not have a significant effect on any adjacent compartments, and thus each compartment could be treated separately from the others.

For each of the nodalised volumes inside containment identified

above, two runs were performed.

In the first of each set of runs, the observer (e.g., equipment item) was, assumed to be on the centerline of the cylinder of concern, midway between the cylinder's ends.

In the second run, the observer was located at the midplane of the cylinder at the inner edge of the compartment.

Although much data was collected from these runs, the results of only two are of particular concezn.

These cases az'e those which give the greatest accumulated doses as a function of time for airborne radiation and for a source in water.

More specifically, the greatest

l I

accumulated doses as a function of time for airborne sources are obtained from case 1 (e.g., using the total free volume of containment after flood-up and the source term given in Table 3-2, with the observer at the center of the cylinder).

Likewise< the greatest accumulated doses as a function of time for submerged sources are obtained from case 4, utilizing the flooded portion of the lower volume, again with.the observer at the centerline of the cylinder.

The accumulated gamma and beta doses for these cases are presented in Tables 3-3 and 3-4g respectively, where the gamma doses given in Roentgen by SHL1GG have been converted to Mrads.

'n order to correct the beta dose values given in Tables 3-3 and 3-4 for attenuation near an equipment item (due to housings,

jackets, etc.), the percentage of accumulated surface beta dose that is attenuated by various thicknesses of three different materials was evaluated.

The result of this evaluation is presented in Tables 3-5 through 3-8.

Zt should be noted that this evaluation's source specific, relying upon the relative activities of the different isotopes and the energies of the various betas emitted.

A plication To The Equi ment uglification Process The results of the calculations described abov'e were applied to electrical components within containment through the following process:

(a) the equipment was identified as either being submerged or in an airborne source; (b) the applicable gross gamma and beta accumulated doses were obtained from Tables 3-3 or 3-4, as appropriate, corresponding to a time equal to or greater than the equipment required operating time; (c) the beta dose was adjusted for shielding effects, if any, utilizing Tables 3-5 through 3-Bg as appropriate>

and (d) the corrected gamma was added to the gross gamma to give the equipment specified post-accident dose.

The following provides examples of how this process was applied to a few categories of equipment:

a)

Category "A".

Cables and cable terminations above flood level were assumed at the center of the total free volume after containment flood-up.

The combined thickness of cable insulation and jacket is conservatively deemed to be 70 mils of unit density material.

Splicing material which protects the cable terminations is also in excess of 70 mils of unit density material.

Credit is taken, however, for only 70 mils of this material.

This category covers Unit 1 System Component Evaluation Worksheets (SCEW sheets, as presented in Attachments 4 and 5 to our letter No. AEP:NRC:0578B dated June ll, 1982)

CC 1 CC 5g CC 6g CC 7g CC Bg CI 1 CZ 2) CI 3p CZ-5( CZ-B, CI-9( CP-9( CP-ll) CP-12(

CP-13(

TC-7) TC-8( TZ-1)

TI 4~ and TP 2~ and Unit 2 SCEW sheets CC lg CC 2g CC 3g CC 4g CC 5~

CC 6~

CC 7~

CC Bg CI 5g CI 7g CI Bg CI 9g CZ-ll, CP-4g CP-5, CP-ll, TC-7, TC-B, TZ-1, TI-2, TI-4, and TP-2.

~

~

~

3-5 b)

Category "B".

Same as in category "A", except that component will be submerged post-accident in flood-up tubes (Kapton insulated wire) which have a wall thickness of 16 mil of steel.

Use Table 3-4 to obtain gross gamma and beta doses to the equipment.

This category covers Unit 1 SCEW sheets CI-ll and CI-12, and Unit 2 SCEW sheets CI-15 and CX-16.

c)

Category "C".

Similar to Category "A", except that the sum of cable insulation and jacket material is less than 70 mils.

This includes Unit 1 SCEW sheets CP>>4 (63 mils), CP-7 (45 mils), and CP-8 (60 mils}, and Unit 2 SCEW sheets CP-6 (63 mils), CP-9 (45 mils), and CP-10 (60 mils).

d)

Category "D".

Components are above flood level and are enclosed in steel enclosures greater *then 70 mils thick which do not permit free air circulation.

Use Table '3-3 and Table 3-6, assuming only 70 mils of steel.

This includes Unit 1 SCEW sheets F-l, LS-l, TC-2, TC-4, TC-16, TC-17, TP-3, V-2, and V-4, and Unit 2 SCEW sheets F-l, LS-l, TC-2, TC-4, TC-16, TP-3, V-2, and V-4.

e)

Category "E".

The electric hydrogen recombiner is above flood level, and beta attenuation is attributed to 10 mils of steel enclosure.

SCEW sheet H-1 (both Unit 1 and Unit 2).

f)

Category "F".

Cable terminations, inside stainless steel flood-up tubing (16 mils wall thickness) will be submerged and are assumed at the center of the lower volume.

No credit is taken for splicing material which protects the terminationsy however, credit for 10 mils of stainless steel is taken.

This category includes SCEW sheets TC-6, TI-3, and TP-1 (both Unit 1 and Unit 2).

g)

Category "G".

Submerged components, assumed at the center of the flooded lower volume.

No credit is taken for beta attenuation due to valve enclosures because they are not assumed to be watertight.

This category includes SCEW sheets EP-01, EP-02, TC-1, TC-3, TI-5, TI-8, TP-S, V-l, and V-5 (both Unit 1 and Unit 2).

Tables 3-9 and 3-10 present, for the Donald C.

Cook Nuclear Plant Unit Nos.

1 and 2, respectively, the results of the calculations for the specific SCEN sheets noted above in categories "A" through "G".

Xt is noted that Unit 1 devices CX-1, CX-2, TC-2, and V-4, and Unit 2 devices TC-2 and V-4 are assumed to have an effective beta attenuation factor of 90%.

This factor is considered to be conservative since these devices are protected by either unit density material or steel in excess of 70 mils thickness.

Table 3-1 Core Isotope Activities

~IOOtO 8

Kr-85 Kr-85m Kr-87 Kr-88 Sr-89 Sr-90 Y-90 Y-91 Zr-95" Zr-97 Nb-95 Nb-95m Nb-97 Mo-99 Tc-99m RQ-103 Ru-106 Rh-103m Rh-105 Rh-106 Ag-110m Ag-ill Sb-125 Sb-127 Te-127 Te-129 Te-129m 8.860 4.300 7.790 1.060 8.060 7.270 7.710 1.060 1.590 1.590 1.590 1.950 1.680 1.770 1.590 1.680 5.850 1.680 1.060 5.850 6.380 5.670 9.750 1.060 1.060 3.280 8.860 x 107 5

x 107 x 10 x 107 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 10 x 108 x 10 x 107 x 105 x 10 x 10 x 107 x 107 x 107 x 10 x 10 XOOtOPO Te-132 I-131 I-132 I-133 I-134 I-135 Xe-133 Xe-133m Xe-135 Xe-135m Cs>>134 Cs-136 Cs-137 Ba-137m Ba-140 La-140 Ce-141 Ce-143 Ce-144 Pr-143 Pr-144 Nd-147 Pm-148 Pm-148m Pm-149 Sm-153 Eu-156 1.420 x 10 8 9.750 x 108 1.420 x 108 1.950 x 108 2.190 x 10 1.700 x 108 1.950 x 107 2.840 x 107 5.310 x 107 5.220 x 10 2.300 x 106 6.380 x 107 1.060 x 106 9.750 x 108 1.680 x 108 1.770 x 108 1.590 x 108 8

1.240 x 108 1.420 x 108 1.240 x 107 6.290 x 10 2'40 x 106 8.860 x 107 6.290 x 107 6.290 x 107 3.280 x 10 Activities of Kr-85m, Kr-87, Kr-88, I-134, I-135, Xe-135, and Xe-135m were obtained from FSAR Table 14.A.2-1 (see July 1982 Updated PSAR).

Other activities were estimated from Table 5-29 of the Westinghouse Standard Information Package, "Radiation Analysis Manual!

Standard Plant Model 412," Revision 3, dated November 1978.

All values are rounded off.

p a

Table 3-2 Case Sl Isotope Activities Total Free Volume Observer At Center

~Zsoto e

Kr-85 Kr-85m Kr-87 Kr-88 Sr-89 Sr-90 Y-90 Y-91 Zr-95 Zr-97 Nb-95 Nb-95m Nb-97 Mo-99 Tc-99m Ru-103 Ru-106 Rh-103m Rh-105 Rh-106 Ag-110m Ag-ill Sb-125 Sb-127 Te-127 Te-129 Te-129m 8.860 x 107 5

4.300 x 107 7.790 x 10 1.060 x 105 8.060 x 104 7.270 x 104 7.710 x 10 1.060 x 10 1.590 x 10 1.590 x 106 1.590 x 10 1.950 x 106 1.680 x 10 1.770 x 10 1.590 x 106 1.680 x 10 5.850 x 10 1.680 x 10 1.060 x 105 5.850 x 10 6.380 x 104 5.670 x 103 9.750 x 10 1.060 x 105 1.060 x 10 3.280 x 104 8.860 x 10

~Isoto e

Te-132 I"131 I-132 I-133 I-134 I-135 Xe-133 Xe-133m Xe-135 Xe-135m Cs-134 Cs-136 Cs-137 Ba-137m Ba-140 La-140 Ce-141 Ce-143 Ce-'144 Pr-143 Pr-144 Nd-147 Pm-148 Pm-148m Pm-149 Sm-153 Eu-156 1.420 x 10 6 4.875 x 10 7.100 x 107 9.750 x 10 1.095 x 107 8.500 x 108 1.950 x 107 2.840 x 10 5.310 x 107 5.220 x 105 2.300 x 104 6.380 x 105 1.060 x 104 9.750 x 106 1.680 x 106 1.770 x 10 1.590 x 106 1.420 x 10 1.240 x 106 1.420 x 106 1.240 x 105 6.290 x 105 2.040 x 104 8.860 x 105 6.290 x 10 6.290 x 105 3.280 x 10

• The activity values listed in this Table are equivalent the core inventory of noble gases, 50% of the core inventory radioiodines, and 1% of the core inventory of particulates.

activity upon release to containment free volume is 1.0012 x Activity concentration upon release to containment is 865.04 of air.

to 100% of of To)al 10 Ci.

Ci per ft

I

~

~

'able 3-3 Case Nl Integrated Doses Total Free Volume Observer At Center Time 0.0 sec Integrated Gamma Dose (Nrads) 0.00 Integrated Beta Dose (Mrads) 0.0 1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> 12.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> 1.0 day 2.0 day 200.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> 400.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> 30.0 day 60.0 day 90.0 day 120.0 day 6.0 month 1.0 year 4.31 20.98 27.96 36.37 61.11 76.50 88.31 99.60 105.76 110.08 116.12 125.80 16.0 87.1 126.3 178.8 338.5 432.7 497.8 553.8 583.5 604.1 632.0 670.8

I

Table 3-4 Case N4 Integrated Doses Submerged Lover Volume -- Observer At Center Time 0.0 sec Integrated'amma Dose (Mrads) 0.00 Integrated Beta Dose (Mrads) 0.0 1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> 12.0 'hour 1.0 day 2.0 day 200.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> 400.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> 30.0 day 60.0 day 90.0 day 120.0 day 6.0 month 1.0 year 0.67 3.44 F 80 6.44 11.2 14.4 17 ~ 2 19.9 21.4 22.4 23.8 26 F 1 0.1 0.8 1.2 1.7 2 ~ 9 3.8 4.5 5.0 5.2 5.3 5.4 5.8

1

Table 3>>5 Surface Beta Attenuation Factors For Unit Density Material Assumin Airborne Source Term Time Thickness of Unit 10 20 30 Density 40 Material (mils) 50 60 70 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 12 hours 1 day 2 days 1 month 2 months 3 months 4 months 4 year 1 year 50.74 35.97 27.52 22.00 18.11 15.21 12.95 35.56 21.57 14.69 10.72 6.96 5.48 4.46 31.92 18.52 12.09 8.50 5.39 4.13 3.29 28.08 15.74 9.95 6.78 4.25 3.18 2.48 18.19 9.51 5.82 3.83 2.42 lo73 1.28 19.26 10.26 6.31 4.14 2.62 1.84 1.34 19.99 10.79 6.66 4.36 2.77 1 ~ 93 1.39 20.42 11.11 6.88 4.50 2.86 1.98 1.42 20.80 11.40 7.07 4.62 2.94 2.03 1.44 20.66 11.33 7.03 4.58 2'2 2 ~ Ol 1.42

• Note:

Factors given are the fractions of accumulated beta dose remaining after a given time.

I II

Table 3-6 k

Surface Beta Attenuation Factors For Stainless Steel Assuming Airborne Source Term Time 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 12 hours 1 day 2 days 1 month 2 months 3 months 4 months 4 year 1 year Thickness of Stainless Steel (mils) 10 20 30 40 50 60 70

1. 11 0.21

0. 08

0. 03

0. 01

0. 01

0. 00 1.12 0.21 0.07 0.03 0.01 0.01 0.00 110 020 007 003 001 001 000 11.47 4.07 1.59 0.64 0.27 0.11 0.05 3.83 1.27 0.51 0.21 0.09 0.04 0.02 2.78 0.89 0.35 0.15 0.06 0.03 0.01 2.08 0.63 0.25 Oelo 0.04 0.02 0.01 1.03 0'5 0.09 0.04 0.02 0.01 0.00 1.06 0.23 0.08 0.03 0.01 0.01 0.00 1.09 0.22 0.08 0.03 0.01 0.01 0.00

• Note:

Factors given are the fractions of accumulated beta dose

~remainin after a given time.

I

'i

Table 3-7 k

Surface Beta Attenuation Factors For Aluminum Assuming Airborne Source Term Time 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 12 hours 1 day 2 days 1 month 2 months 3 months 4 months 4 year 1 year 10 Thickness of 20 30 Aluminum (mils) 40 50'0 70 11.31 3.76 1.95 1.22 0.83 0.59 0.42 6.68 2.10 0.95 0.53 0.34 0.23 0.16 7.24 2.27 0.98 0.52 0.32 0.21 0.15 7.64 2.39 1.00 0.52 0.31 0.20 0.14 7.88 2.46 1 ~ 01 0.51 0.30 0.20 0.13 8.10 2.53 1.02 0.51 0.29 Oe19 0.13 8.05 2.51 1.00 0.49 0.28 0.18 0.12 29.65 16.86 10.98 7.53 5.30 3.78 2.72 16.33 6.30 3.63 2.39 1.65 1.18 0.85 13.61 4.82 2.63 1.69 1.16 0.82 0.59

• Note:

Factors given are the fractions of accumulated beta dose

~remainin after a given time.

v 1

( ~

Table 3-8 Surface Beta Attenuation Factors For 10 Mils Of Stainless Steel Assumin Submergence Source Term

. Time 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Percentage Of Beta Dose Remaining F 88 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1 day 2 days 1 month 2 months 3 months 4 months 4 year 1 year

l. 50 1.19 0.98 0.57 0.56 0.55 0.54 0.52 0.49

Table 3-9 Donald C. Cook Nuclear Plant Unit No.

1 S ecified Radiation Doses Inside Containment Device CC-1 CC-5 CC-6 CC-7 CC-8 CI-1 CZ-2 CI>>3 CI-5 CZ-8 CZ-9 CZ-11 CZ-12 CP-4 CP-7 CP-8 CP-9 CP-11 CP-12 CP-13 EP-01 EP-02 P;1 H-1 LS-1 TC-1 TC-2 Equi ment Descri tion Continental Cable 3119 Continental Cable 3121 General Electric 3121 Continental Cable 3122 General Electric 3121 Boston Ins'd. Wire 3064 Rockbestos 3064 Samuel Moore 3075 Boston Ins'd. Wire 3075 Cerro 3077 Samuel Moore 3077 Kapton Insulated Wire Kapton Insulated Wire Okonite 399 Anaconda 3116 Essex 3116 Kerite 3116 Kerite 3127 Cyprus 347 Anaconda 347 4 kV Penetrations Penetrations Pan Motors Hydrogen Recombiner NAMCO Limit Switch Control Cable Term.

Control Cable Term.

Specified Operating Time 14 days 14 days 14 days 14 days 14 days 5 seconds 5 seconds 4 months 4 months 4 months 4 months 1 year 10 seconds 3 months 14 days 14 days 14 days 1 year 3 months 3 months 1 year 1 year 1 year 3 months 14 days 30 minutes 30 minutes Gamma Dose, (MRads) 76.50 76.50 76.50 76.50 76.50 4.31 4.31 110.08 110.08 110.08 110.08 26.1 0.67 105.76 76.50 76.50 76.50 125.80 105.76 105.76 26.1 26.1 125.80 105.76 76.50 0.67 4'1 Beta Dose,

Gross, (MRads) 432. 7 432.7 432. 7 432.7 432.7 16.0 16.0 604.1 604.1 604.1 604.1 5.8 0.1 583.5 432.7 432.7 432.7 670.8 583.5 583.5 5.8 5.8 670.8 583.5 432.7 0.1 16.0 Beta Dose, After Attenuation, (MRads}

10.73 10.73 10.73 10.73 10.73 1.6 1.6 8.58 8.58 8.58 8.58 0.03 0.1 11.26 29.34 13.76 10.73 9.53 8.11 8.11 5.8 5.8 0.00 6.36 0.04 0.1 1.6 Effective Specified

Dose, (MRads) 87.23 87.23 87.23 87.23 87.23 5.91 5.91 118.66 118-66 118.66 118.66 26.13 0.68 117.02 105.84 90.26 87.23 135.33 113.87 113.87 31.9 31.9

,125-80 112.12 76.54 0.77 5.91

Table 3-9 (continued)

Device Equi ment Descri tion Specified Operating Time Gamma Dose, (MRads)

Beta Dose,

Gross, (MRads)

Beta Dose, Effective After Specified Attenuation,

Dose, (MRads)

(MRads)

TC-3 TC-4 TC-6 TC-7 TC-8 TC-16 TC-17 TI-1 TI-3 TI-4 TI-5 TI-8 TP-1 TP-2 TP-3 TP-5 V-1 V-2 V-4 V-5 Control Cable Term.

Control Cable Term.

Control Cable Term.

Control Cable Term.

Cable Termination Control Cable Term.

Seal Assembly Instrument Cable Term.

Instrument Cable Term.

Instrument Cable Term.

Instrumentation Term.

Instr. Cable Term.

Power Cable Term.

Termination Termination Power Cable Term.

Valve Motor Operator Valve Motor Operator Valve Motor Operator Valve Motor Operator 30 minutes 1 day 14 days 14 days 14 days 1 day 14 days 4 months 4 months 4 months 5 minutes 32 seconds 1 year 1 year 1 year 30 minutes 30 minutes 1 day 30 minutes 60 seconds 0.67 27.96 14.4 76.50 76.50 27.96

.76.50 110.08 22.4 110.08 0.67 0.67 26.1 125.80 125.80 0.67 0.67 27.96 4.31 0.67 0.1 126.3 3.8 432.7 432.7 126.3 432e7 604.1 5.3 604.1 0.1 0.1 5.8 670.8 670.8 0.1 0.1 126.3 16.0 0.1 0.1 0.01 0.04 10.73 10.73 0.01 0.04 8.58 0.03 8.58 0.1 0.1 0.00 9.53 0.00 0.1 0.1

0. 01 1.6 0.1 0.77 27.97 14.44 87.23 87.23 27.97 76.54 118.66 22.43 118.66 0.77 0.77 26.1 135.33 125.80 0.77 0.77 27.97 5.91 0.77

Table 3-10 Donald C. Cook Nuclear-Plant Unit No.

2 S ecified Radiation Doses Inside Containment Device Equipment Description Specified Operating Time Beta Dose, Gamma Dose,

Gross, (MRads)

(MRads)

Beta Dose, After Attenuation, (MRads)

Effective Specified

Dose, (MRads)

CC-1 CC-2 CC-3 CC-4 CC-5 CC-6 CC-7 CC-8 CZ-5 CZ-7 CZ-8 CI-9 CI-3.3.

CI-15 CI-16 CP-4 CP-5 CP-6 CP-9 CP-10 CP-11 EP-01 EP-02 F-1 H-1 LS-1 Continental Cable 3119 Continental Cable 3120 General Electric 3120 Anaconda 3120 Continental Cable 3121 General Electric 3121 Continental Cable 3122 General Electric 3122 Samuel Moore 3075 Boston Ins'd. Wire 3075 Cerro 3077 Samuel Moore 3077 Boston Ins'd. Wire 3077 Kapton Insulated Wire Kapton'nsulated Wire Cyprus 347 Anaconda 347 Okonite 399 Anaconda 3116 Essex 3116 Kerite 3116 4 kV Penetrations Penetrations Fan Motors Hydrogen Recombiner NAMCO Limit Switch 14 days 14 days 14 days 14 days 14 days 14 days 1 day 1 day 4 months 4 months 4 months 4 months 4 months 1 year 10 seconds 1 year 1 year 3 months 14 days 14 days 14 days 1 year 1 year 1 year 3 months 14 days 76.50 76.50 76.50 76.50 76.50 76.50 27.96 27.96 110.08 110.08 110.08 110.08 110.08 26.1 0.67 125.80 125.80 105.76 76.50 76.50

76. 50 26.1 26.1 125.80 105.76

76. 50 432. 7 432. 7 432. 7 432. 7 432. 7 432.7 126.3 126.3 604.1 604.1 604.1 604.1 604.1 5.8 0.1 670.8 670.8 583.5 432.7 432.7 432.7 5.8 5.8 670.8 583.5 432.7 10.73 10.73 10.73 10.73 10.73 10.73 4.16 4.16 8.58 8.58 8.58 8.58 8.58 0.03 0.1

9. 53 9.53 11.26 29.34 13.76 10.73 5.8 5.8 0.00 6.36 0.04

87. 23 87.23 87.23 87.23 87.23 87.23 32.12 32.12 118.66 118.66 118.66 118.66 118.66 26.13 0.68 135.33 135.33 117.02 105.84 90.26 87.23 31-9 31.9 125.80 112.12 76.54

Table 3-10 (continued)

Device Equi ment Descri tion Specified Operating Time Beta Dose, Gamma Dose,

Gross, (MRads)

(MRads)

Beta Dose, After Attenuation, (MRads)

Effective Specified

Dose, (MRads)

TC-1 TC-2 TC-3 TC-4 TC-6 TC-7 TC-8 TC-16 TI-1 TI-2 TI-3 TI-4 TI-5 TI-8 TP-1 TP-2 TP-3 TP-5 V-1 V-2 V-4 V-5 Control Cable Term.

Control Cable Term.

Control Cable Term.

Control Cable Term.

Control Cable Term.

Contxol Cable Term.

Control Cable Term.

Contxol Cable Term.

'Instrument Cable Term.

Instrument Cable Term.

Instrument Cable Term.

Instrument Cable Term.

Instrumentation Term.

Instr. Cable Term.

Power Cable Term.

Power Cable Term.

Termination Power Cable Term.

Valve Motor Operator Valve Motor Operator Valve Motor Operator Valve Motor Operator 30 minutes 30 minutes 30 minutes 1 day 14 days 14 days 14 days 1 day 4 months 4 months 4 months 4 months 5 minutes 32 seconds 1 year 1 year 1 year 30 minutes 30 minutes 1 day 30 minutes 60 seconds 0.67 4-31 0.67 27.96 14.4 76.50 76.50 27.96 110.08 110.08 22.4 110.08 0.67 0.67 26.1 125.80 125.80 0.67 0.67 27.96 4'1 0.67 0.1 16.0 0.1 126.3 3.8 432.7 432.7 126.3 604 '

604.1 5.3 604.1 O.l 0.1 5.8 670.8 670.8 0.1 0.1 126.3 16.0 O.l 0.1 1.6 0.1 0.01 0.04~

10.73 10.73 0.01 8.58 8.58 0.03 8.58 0.1 0.1 0.00 9.53 0.00 0.1 0.1 0.01 1.6 0.1 0.77 5.91 0.77 27.97 14.44 87.23 87.23 27.97 118.66 118.66 22.43 118.66 0.77 0.77 26.1 135.33 125.80 0.77 0.77 27.97 5.91 0.77

ATTACHMENT 4 TO AEP:NRC:0578I DESCRIPTION OF 'CALCULATIONALPROCEDURE USED IN DETERMINING RADIATION DOSES TO ELECTRIC E UIPMENT OUTSIDE CONTAINMENT

'I 0

4-2 Introduction In order to obtain bounding estimates for radiation doses to equipment located near pipes which carry fluid recirculated from the containment (for the purpose of long-term'post-LOCA heat removal),

radiation calculations were performed for a number of piping systems outside containment.

The results of these calculations and their application to the equipment qualification process are described in this attachment.

Piping Calculations The original piping calculations of concern were performed in April of 1980.

The calculations were performed using a computer program entitled "NSLSHL3" which modeled the source in a cylindrical pipe containing radioactive matex'ial as a set of line sources.

This code is, in effect, an earlier version of the SHLlGG code described in Attachment 2 to this submittal.

e The input data required included parameters to characterize the pipe material and dimensions, the target (i.e., observer or equipment) location, and the activities, gamma energies, and decay rates of the radioactive isotopes in the fluid.

Because the calculations were concerned only with targets outside the pipe, beta radiation dose was assumed to be negligible when compared with the associated gamma dose.

Furthermore, the only output of concern was assumed to be accumulated (integrated) gamma dose as a function of time.

For all calculated

cases, the source in the pipe was assumed to be distributed in water; air was assumed to surround each pipe.

All pipes were assumed to be made of iron, with no additional shielding present.

In sum, eleven (ll) dose calculations for the pipes were made.

For cases 1 through ll, the volume of water into which the source was distributed was assumed to be equal to the minimum volume available for recirculation at the time of;switchover from the injection to the recirculation phase of a LOCA.

This volume was defined as consi.sting of the difference between the minimum Refueling Water Storage Tank (RWST) content (350,000 gallons) and the RWST low level set point (131,962 gallons), plus the Reactox'oolant System inventory of 93,960 gallons.

This total water volume is, therefore, 311,998 gallons.

For cases 1 through 10, the fraction of core radioactive inventory in the water was assumed to be 100% of the noble gases, 50% of the radioiodines, and 1% of the other fission products.

Since it was deemed highly unlikely that the noble gases would actually be in the recix'culating fluid (rather, they should remain in the containment atmosphere),

case ll was performed with no noble gases present.

Since

4 3

the input for case 4 and case 11 were identical with the exception of the assumed noble gas source, a base was provided to assess the effect of the noble gas sgurce term.

Indeed, the one>>year accumulated gamma doses of 10.2 x 10 R and 8.92 x 10 R for cases 4 and ll, respectively, indicated that inclusion of the noble gases in the calculations could provide a conservative margin on the order of 14% to the estimated doses.

The piping systems considered in the calculations involved portions of the Centrifugal Charging (CC), Safety Injection (SI), Residual Heat Removal (RHR), and Containment Spray (CS) systems.

These systems are located outside containment and may be expected to carry fluid recirculated from the containment sump.

The specific pipe sizes considered are presented in Table 4-1.

Bounding cases were run for the largest diameter pipe (14.0" outside diameter) and for the pipes with the largest outside diameter-to-wall thickness ratios (OD/T = 58.28; OD/T = 49.44).

All targets were assumed at a location equidistant from the ends of the length of pipe considered.

Cases 1 through 4 were run to test the sensitivity of the results to the choice of radial and azimuthal divisions of the source.

A matrix of 20 radial and 36 azimuthal divisions was found to be satisfactory for even the largest diameter pipe.

This method of division was therefore used in cases 4 through 11.

Cases 4 through 6 were run to compare piping doses as a function of OD/T.

The dose was found to increase as OD/T increased, even though the pipe ODs varied.

Apparently the pipe sizes considered were not of prime importance; it appeared as though, for the specific selection of pipes, the most important factor was whether the pipe wall was thin enough to minimize gamma attenuation.

Cases 4, 7, and 8 were run to test the effect of locating the target at various distances from the 14" OD pipe.

In particular, these cases assumed the target at the pipe outer surface, 13" from the pipe centerline, and 31" from the pipe centerline.

The resultant one-year accumulated doses expressed in MR for these cases were, respectively, 1.02, 0.52, and 0.21.

Cases 4, 6, 9, and 10 were run to assess the effect of an assumed pipe length on the calculations.

There was no difference in integrated doses due toa 12.5 ft or a 25.0 ft length of pipe (assuming the target at the pipe outer surface).

In summation, the bounding case found as a result of this work was case 6.

This case assumed 100% of the core inventory of noble gases, 50% of the core inventory of radioiodines, and 1% of the core inventory of other fission products to be distributed in approximately 312<000 gallons of water.

The assumed pipe dimensions were 8.625" OD, 0.148" wall thickness, and 12.5 ft length.

,Utilizing 20 radial and

$6 azimuthal divisions, a one-year integrated dose of 14.58 x 10 R was calculated at a point on the pipe outer surface, midway between the pipe ends.

Multiplying this value by 1.3 to account for daughter sources and

4>>4 by 0.875 tg convert to rads yields a one-year accumulated dose of 16.58 x 10 rads for this bounding case.

(his value compares favorably to the one-year integrated dose of 12 x 10 rads specified for equipment outside containment (not subject to submersion) in information transmitted to us by Westinghouse Electric Corporation.

The integrated dose at various times up to one year for case 6 were also determined to be as follows:

at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 1.02 Mrads; at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 2.62 Mrads; at 1

day, 3.98 Mrads; at 1 month, 12.3 Mrads; at 4 months, 14.79 Mrads; and at one year, 16.58 Mrads.

A lication To The Equi ment Qualification Process Based upon the results presented above for case 6, the accumulated gamma dose as a function of time up to one year was applied to various equipment items located outside containment.

As an example, if an equipment item had a required operating time of one month, the accumulated gamma dose corresponding to that time frame as determined in case 6 (e.g.,

12.3 Mrads) was applied to the item. If the equipment item was then determined to be located at a distance from the nearest

pipe, a factor applicable to that distance (i.e., 1.0 for equipment at the pipe outer surface, 0.51 for equipment 6" from the pipe outer

surface, and 0.21 for equipment, 2 ft from the pipe outer surface) would then be multiplied by the dose determined from case 6 to yield the equipment specified post-accident.

dose.

These correction factors were obtained by comparing the one-year accumulated doses for cases 4, 7, and 8 (see discussion on calculation findings above).

Using this calculational procedure and plant specific installation

records, specified doses were obtained for most of the electric equipment of concern outside containment.

In general, these equipment items were treated as belonging to any of the four following categories:

a)

Cables and cable terminations at instruments and solenoids which are not subject to radiation are treated as category "A"

items.

Such items include those shown on Unit 1 SCEW sheets CI-4, CI-10, TC-15'I-9, and TI-10( and Unit 2 SCEH sheets CI-6, CI-10, TC-15, TI-9, and TI-10.

b)

Components assumed at, the surface of the pipe are category "B"

items.

Such items include those on Unit 1 SCEW sheets CC-2, CC 3J CC 4g CC 9J CC lOJ TC lOJ TC llew TC 13'P 6~

V 6~

V 7g and V-ll, and Unit 2 SCEW sheets CC-9, CI-12, CI-14, TC-10>

TC 11'C 13>

TP 6g V 6 V 7p and V ll c)

Components in category "C" are those located at distances greater than or equal to 6" from a recirculation line, but which are conservatively assumed to be at a 6" distance.

This applies to Unit 1 SCEW sheets CP-l, CP-3, CP-5, CP-10, M-1g M>>2, and TP-4, and Unit 2 SCEW sheets CP<<l, CP-2, CP>>8, CP-13, M<<l, M-2, and TP-4.

4-5 d)

Components in category "D" are those located at distances greater than or equal to 2 ft from a recirculation line, but which are conservatively assumed to be at a 2 ft distance.

This aPPlies to Unit 1 SCEW sheets CP-6, TC-14, and V-10, and Unit 2 SCEW sheets CP-7, CP-12, TC-14, and V-10.

Specified radiation doses for these equipment items in categories "A" through "D" are presented in Tables 4-2 (Unit 1) and 4-3 (Unit 2).

t.

Table 4-1 Pipes Considered In Calculating Outside Containment Doses

~Sstem CC CC CC CC CC SI SI SI SI SI SI CS CS CS CS CS CS CS CS M-14 M-14 M-14 B-14 B-14 K-14 B-14 B-14 K-14 K-14 K-14 G-14 G-14 G-14 E-14 E<<14 E-14 E-14 E-14 B-14 G-14 G-14 Outside Diameter, OD (inches) 4.5 3.5 2.0 6.625 8.625 4.5 4.5 6.625 2.375 1.9 1.05 14.0 3.5 8.625 10.75 8.625 6.625 3.5 1.315 8.625 12.75

2. 375 Wall Thickness, T (inches) 0.438 0.438 0.344 0.134 0.148 0.337 0.12

0. 134 0.276 0.2 0.154 0.438 0.216 0.322 Oe365 0.322 0.28

'0. 216 0.133 0.148 0.406 0.154 OD/T 10.3 7.99 5.81 49.44 58.28 13.35 37.5 49 ~ 44 8.61 9.5 6.82 31.96 16.20 26.79 29.45 26.79 23.66 16.20 9.89 58.28 31.40 15.42

• Note:

Pipe designation indicates materials of construction, seismic classification, and quality level.

For further information, AEPSC specifications should be consulted.

Table 4-2 Donald C. Cook Nuclear Plant Unit No.

1 S ecified Radiation Doses Outside Containment Device Equipment Description Specified Operating Time Specified Rads (x 10

)

CC-2 CC-3 CC-4 CC-9*

CC"10 CI-4 CI-10 CP-1 CP-3 CP-5 CP-6 CP-10 M-1 M-2 TC-10 TC-11 TC-13 TC-14 TC-15 TI-9 TI 10 TP-4 TP-6 V-6 V-7 V-10 V-11 Continental Cable 3120 General Electric Cable 3120 Anaconda Cable 3120 Continental Cable 3123 General Electric Cable 3123 Continental Cable 3075 Continental Cable 3077 Okonite Cable 324 Essex Cable 324 Anaconda Cable 3102 Okonite Cable 3102 Anaconda Cable 3103 CVCS p SI E RHR (Pump Motors)

Containment Spray (Pump Motors)

Term. At Valve Motor Operator Term. At Valve Motor Operator Termination At Terminal Block Term. At Valve Motor Operator Termination At Solenoid Term. At Barton Instruments Term. At Foxboro Instruments Termination At Pump Motor Termination At Valve Motor Valve'otor Operator Valve Motor Operator Valve Motor Operator Valve Motor Operator 1 day 1 day 1 day 1 day 1 day 4 months 4 months 1 month 1 month 1 year 1 month 1 month 1 year 1 month 1 day 1 day 1 day 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 25 seconds 4 months 4 months 1 year 1 day 1 day 1 day 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1 day 3'8 3.98 3'8 3.98 3.98 Not Applicable Not Applicable 6.40 6.40 8.62 2.58 6.40 8.62 6.40 3.98 3e98 3'8 0.21 Not Applicable Not Applicable Not Applicable 8.62 3'8 3.98 3.98 0.21 3'8

Table 4-3 Donald C.

Cook Nuclear Plant Unit No.

2 S ecified Radiation Doses Outside Containment Device D ci ment Deecti ticn Specified Operating Time Specified Rads (x 10

)

CC-9 CI-6 CI-10 CI-12 CI-14 CP-1 CP-2 CP-7 CP-8 CP-12 CP-13 M-1 M-2 TC-10 Tc<<ll TC-13 TC-14 TC-15 TI-9 TI-10 TP-4 TP-6 V-6 V-7 V-10 V-ll Continental Cable 3123 Continental Cable 3075 Continental Cable 3077 Raychem Cable 3111 Continental Cable 3069 Essex Cable 324 Cyprus Cable 324 Cyprus Cable 3102 Okonite Cable 3102 Anaconda Cable 3102 Anaconda Cable 3103 CVCS, SI, RHR (Pump Motors)

Containment Spray (Pump Motors)

Term. At Valve Motor Operator Term. At Valve Motor Operator Termination At Terminal Block Term. At Valve Limit Switch Termination At Solenoid Term. At Barton Instruments Term. At Foxboro Instruments Termination At Pump Motor Termination At Valve Motor Valve Motor Operator Valve Motor Operator Valve Motor Operator Valve Motor Operator 1 day 4 months 4 months 1 day 1 day 1 month 1 month 1 month 1 year 1 month 1 month 1 year 1 month 1 day 1 day 1 day 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5 seconds 4 months 4 months 1 year 1 day 1 day 1 day 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1 day 3.98 Not Applicable Not Applicable 3'8 3'8 6.40 6.40 2.58 8.62 2.58 6.40 8'2 6.40 3.98 3.98 3'8 0.21 Not Applicable Not Applicable Not Applicable 8 ~ 62 3 ~ 98

3. 98

3. 98 0.21

3. 98

A

~ 1

~

0