ML19331D722

Forwards Responses to NRC 800807 Questions Re Shielding of Personnel After Accident.Responses Are in FSAR Change Form & Will Be Included in Next FSAR Amend

ML19331D722
Person / Time
Site:	Summer
Issue date:	08/27/1980
From:	Nichols T SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8009030493
Download: ML19331D722 (16)

Text

-

SOUTH CAROLINA ELECTRIC a GAS COMPANY

( .ou on.cc .o n.

Cotume A.SouTM CAnouMA 29218 T.C.NiewoLs,Jn.

vc .......u.. o.ov. c cev=. August 27, 1980 (Nuclear Operations)

Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395

^

Radiation Protection Questions -

Sheilding

Dear Mr. Denton:

South Carolina Electric and Gas Company, acting for itself and as agent for South Carolina Public Service Authority, herewith provides forty-five (45) copies of our responses to questions raised in an August 7, 1980 phone call with your Mr. John Minns tegarding shielding of personnel af ter an accident. The responses are in the form of an FSAR change and will be included in the next FSAR amendment.

If you require additional information, please let us know.

Very truly yours,

, & ,//

T. C. Nichols, Jr.

RBC:TCN:rh cc: G. H. Fischer E. H. Crews, Jr.

V. C. Summer W. A. Williams, Jr.

B. A. Bursey D. A. Nauman O. S. Bradham M. A. Barnisin R. B. Clary H. T. Babb

0. W. Dixon PRS / Woods NPCF/Whitaker File 8009080IM3 .. - . -. _- . . .- .. . ... . .. . -

331.34 Provide a su= mary of the shielding design review required by our letter dated November 9, 1979, implementing the Lessons Learned item 2.1.6.b of NUREG-0578, and provide a description of the results of this review.

Include in your description.

a. source terms used in the evaluation (NUREG-0578 specified that source terms in Regulatory Guide 1.3,1.4 and 1.7 be used) .

b. systems assumed to contain high levels of radioactivity in a post-accident situation including, but not limited to, containment residual heat removal, safety injection, CVCS, demineralizers, charging systems, reactor coolant filters, seal water filters, sample lines, liquid radwaste systems, and gaseous radwaste systems. If any of these systems or others that could contain high radioactivity were excluded, explain why such systems were excluded from review. You should verify that direct radiation from field run piping and scattered radiation (such as shine over shield walls) were included in the analysis,

c. specify areas where access is considered necessary for vital system operation after an accident. Your evaluation of areas to determine the necessary vital areas should include but not be limited to, consideration of the control room, Technical Support Center, Operational Support Center, recombiner hookup and control stations, hydrogen purge control stations, containment isolation reset control area, S sampling and sample analysis areas, manual ECCS alignment area, motor control centers, instrument panels, emergency power supplies, security center and radwaste control panels. If any of these areas were not considered areas where access was necessary af ter an accident, explain why they are excluded.

d. Designation of the codes used for analysis, such as ORIGEN, IS0 SHIELD, QUAD or others.

e. The projected doses to individuals for necessary occupancy times in vital areas.

f. A brief description of the ptoposed plant modifications resulting from the design review and confirmation that these modifications will be complete by full power operation.

RESPONSE

Refer to the revised Sections 12.1.1 and 12.1.3, and Appendix 12A. The report prepared for the shielding design re. view that was performed in response to NUREG-0578 has been incorporated into the V. C. Summer FSAR as Appendix 12A. The contents of Appendix 12A provide a comprehensive explanation of the shielding design review.

a. The origin of the source terms used in the NUREG-0578 shielding design review are explained in Section 12A.3-1 of Appendix 12A. The initial isotopic inventory of the shielding source terms are presented in Table 12A.3-2 of Appendix 12A.

b. The following systems either in their entirety or in part are assumed to contain high levels of radioactivity in a post-accident situation:

(1) Residual Heat Removal System (2) Safety Injection System (3) Reactor Buildind Spray System (4) Reactor Coolant System f

(5) Post-accident Hydrogen Removal System i

This system is of concern only for the sections involved in the collection of the reactor building atmosphere sample. -

(6) Nuclear Sampling System This system is of concern for its role in the collection of the .

following samples: primary coolant hot legs, pressurizer liquid and steam spaces, residual heat removal loop.

(7) Chemical and Volume Control System This system will be of concern only for those system sections involving high head injection and seal water injection. The high head injection section involves the charging / safety injection pumps and the associated piping required to charge the reactor coolant system. The pumps will initially receive their suction supply from the refueling water storage tank and then from the reactor building recirculation sump during the recirculation mode. The seal water injection section of the CVCS will be isolated prior to the start of the recirculation mode and will, therefore, not contain the high radioactive levels found in the reactor building recirculation sump.

The CVCS sections involving the letdown and purification functions are not required for the post-accident situation, and will be isolated from the high levels of radioactivity contained in the coolant and sump liquids. The demineralizers and filters of the CVCS have been excluded from the review, since the amount of fuel damage present will render their usage undesirable.

(8) Radwaste Gas Handling Systeta Although not designed for reactor coolant degasification under post-accident conditions, the waste gas system was evaluated for the exposures emitted and received under the post-accident conditions.

The following systems are excluced from the review because they will be isolated upon the occurrence of an accident and will therefore not contain the high radiation levels resulting from the fuel damage:

(1) Radwaste Solids Handling System (2) Radwaste Liquid Handling System (3) Spent Fuel Cooling System (4) Boron Recycle System (5) Thermal Regeneration System The direct radiation dose from all piping of significance to an area has been considered in the calculation of the doses presented in Appendix 12A. Contributions to the dose from scattered radiation were considered in the review during the calculation of both the dose and dose rates for the access areas in the plant.

c. The areas where access is considered necessary for the opt ation of vital systems after an accident are given in Sections 12A.4.1 through 12A.4.19.

For those areas specifically requested for consideration:

e----

3 (1) Control Room = refer to Section 12A.4.15 (2) Technical Support = refer to Section 12A.4.14 Center

! (3) Operational Support = part of the Control Room, refer to Center Section 12A.4.15.

(4) Recombiner hookup and = refer to Sections 12A.4.11 and 12A.4.12 control station (5) Hydrogen purge control = the alternate purge system controls of stations the Post Accident Hydrogen Removal System are located in the Control Room, refer to Section 12.A.4.15.

! (6)' Containment isolation = in the Control Room, refer to Section reset control area 12.A.4.15 (7) Sampling and sample = refer to Sections 12A.t.13, 12A.4.16, ,

analysis areas 12A.4.17, and 12A.4.18 (8) Manual ECCS alignment = refer to Sections 12A.4.3, 12A.4.,7 areas 12A.4.9, and 12A.4.10 l (9) Motor Control Centers = refer to Sections 12A.4.7 and 12A.4.10 (10) Instrument panels = refer to Sections 12A.4.3, 12A.4.5, 12A.4.6, 12A.4.8, 12A.4.11, 12A.4.12, and 12A.4.15 (11) Emergency power supplies = refer to Sections 12A.4.7 and 12A.4.10 (12) Security center = part of Service Building. Not a vital access area.

(13) Radwaste control panels = refer to Sections 12A.4.5 and 12A.4.8

d. Computer codes used in the analysis are INHEC, RWDS, and SCC.

I i (1) INHEC For information on INHEC, refer to the following Topical Report:

l " Computation of Radiological Consequences Using INHEC Computer Program" lj Gilbert / Commonwealth Companies l

Topical Reports GAI-TR-101P-A (Proprietary)

! and GAI-TR-101-A (Non-Proprietary)

March, 1976.

1

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

. . l l

(2) RWDS The .RWDS computer code performs the calculation of the activity levels in both the fluid and components of a system. This code incorporates numerous options which allow for the analysis of various systems with a broad spectrum of characteristics. Using this computer code, it is possible to generate the isotopic inventory of the individual system components, as well as the system flow stream, after the individual isotopic components of the stream have been decayed independently of the others. Capable of handling the activities of several streams simultaneously, the RWDS code can also generate the total activity present when several streams are merged into a single stream. The activities of the inventory are summed by energy group, and by liquid and gaseous activity, so as to generate a input for= from the list of isotopes and activities suitable for use in shieldir.g analysis. The activities can be obtained in ter=s of either specific or total activities present in all the contributing streams.

(3) SDC The SDC computer code is designed to calculate the gamma-ray shielding requirements for nuclear applications. The integration of the basic exponential attenuation point kernel over the varicus source geometries provides an uncollided gam =a-ray flux. The biological dose rate is obtained by multiplying this uncollided flux by the product of a flux-weighted buildup factor and a dose conversion factor. The major options in the computer code permit calculation of either the required shield thickness when a dose rate is specified, or the dose rate when the shield thickness is given. The calculation of dose rates from

unshielded sources as well as surface intensities for cylinders and spheres is also included.

e. Refer to Table 12A.4-2 for the radiation dose to individuals in the vital areas.

f. Plant modifications that =af be required due to the reviev are currently under study and will be provided in a later submittal.

12A.3 SOURCE TERMS AND CALCUIATIJNAL METHODOLOGY 12A.3.1 SOURCE TERMS 12A.3.1.1 Basis for the Source Terms The activity releases assumed in this eview are based on the assumptions and regulatory positions contained in the Regulatory Guides 1.4 and 1.7. The activity assumed for liquid source term calculation is based on 100% of the noble gas inventory, 50% of the halogen core inventory and 1% of all other nuclides in the core inventory. The activity assumed for gaseous source term calculation is based on 100% of the noble gas core inventory and 25% of the halogen core inventory.

12.A.3.1.2 Liquid Source Terms Two liquid sources are considered in the design review: (1) the undiluted fluid cs found within the reactor coolant system, and (2) the diluted fluid as found within the reactor building recirculation sump. The first source term, undiluted reactor coolant, is required in the examination of those systems whose flow originates from either the Reactor Coolant System or an auxiliary system containing the undiluted primary fluid. The source terms are based on the dilution of the liquid activity inventory discussed in the first paragraph with the fluid volume of the Reactor Coolant System. This source is used for the examination of the reactor coolant fluid sampling section of the Nuclear Sampling System, and those portions of the Chemical and Volume Control System associated with the degasific: tion of the reactor coolant fluid.

In the secotd liquid source term, consideration is given for the dilution of the i liquid activity inventory discussed it the first paragraph with the fluid volume l contained in the reactor building recirculation sump. The minimum fluid volume expected in tie sump, and the individual contributors to that volume, are given in Table 12A.3-1. These source terms are utilized in the examination of those systems which receive their fluid supply from the reactor building recirculation sump. Those a ystems which are considered in the review are:

4 Residual Heat Removal System Reactor Building Spray System i

Safety Injection System Nuclear Sampling Syste= (RHR process fluid sample) 12A.3.1.3 Gas Source Ter=s L

The gaseous source terms were detecnined for containment and the waste gas system i using the activity releases described in Section 12A.3.1.1.

l The containment airborne source term was based on the dilution of the gaseous j activity inventory by the containment free volume atmosphere.

Although not designed for reactor coolant degasification under post-accident conditions, the waste gas system was evaluated for the exposures emitted and received under the post-accident conditions. The waste gas system is designed to remove the fission product gases from the reactor coolant contained in the Volume Control Tank (VCT). The amount of fission gases re=oved from the reactor coolant in the VCT and collected by the waste gas system can be related to the amount entering the VCT as follows:

)

(1) Stripping Efficiency (SE):

CR ~ L SE = C -C R L eq (2) Stripping Fraction (SF):

C g -C L

SF =

k where C = the gas concentration in the reactor coolant liquid entering the R

Volu=e Control Tank, i

a-f a

d

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

Cg = the gas concentration in the reactor coolant liquid leaving the Volume Control Tank, Cg = the gas concentration in the reactor coolant liquid leaving the

• 9 Volume Control Tank, assuming the ratio of the gas concentration in the liquid and gas phases in the Volume Control Tank follows Henry's Law.

The waste gas system source terms were determined for the degasification of the reactor coolant liquid by the calculation of the quantity of activity entering the Volume Control Tank via tha normal letdown path. For the sake of conservatism, the stripping efficiency of this process is assumed to be 100%. Therefore, in the previous equation of the stripping fraction, C =C g , and the stripping fraction is: "9 CR ~ L SF =

CR Thus, the separation of the fission gases from the reactor coolant liquid in the VCT will follow Henry's Law. This results in the maximum theoretical gas concentration in the vapor phase of the VCT and, hence, the maximum quantity of '

gas enters the waste gas system.

12A.3.2 METHODOLOGY 12A.3.2.1 Calculation of Dose Rates Dose rates for the areas of interest in this review were calculated by determining the potential contributing sources at a representative location and using the appropirate source term from Table 12A.3-2 adjusted for decay as required. The dose rate at the representative location was used as the general area dose rate for the area. The SDC computer code (Ref. 1) was used in performing the dose rate claculations. Energy groups required as input to the computer code ware determined using the gamma ray energy and intensity data in Refs. 2 and 3 for the nuclides in Table 12A.3-2. j

y__

i 12A.3.2.2 Calculation of Doses to Personnel During Post Accident Access to Vital Areas Personnel doses received in performing a specific task in a given vital area are calculated as the sum of the doses received during travel to and from the vital area and the dose received while performing the given operation in the vital area.

The doses received during travel are determined by calculating dose rates at selected locations (or at a single location if the dose rate along the travel route is relatively uniform) along the travel route using the methodology i discussed in Section 12A.3.2.1 and multiplying the dose rates by the appropriate travel time for each selected location along the travel route.

Doses received while performing a given operation are determined by multiplying i

the dose rate for the given area by the time required to perform the operation.

Dose rates for the given vital area are determined using the methodology discussed in Section 12A.3.2.1.

12A.3.2.3 Calculation of Integrated Doses to Safety Equipment The integrated dose to a given ites of safety equipment is determined by integrating the dose rate appropriate for the given ites over the time period that it is required to be available to perform its safety function. - Dose rates are calculated using the methodology discussed in Section 12A.3.2.1.

s

_. __ _ - _ _ , _- -_. - -- . _ . _ _ __, ~. -

p-TABLE 12A.3-1 Containment Sump Minimum Liquid leventory Liquid Source Liquid Volume (ft )

I l

Refueling Water Storage Tank 46,791* t 4

Safety Injection Accumulators 3,000 Boron Injection Tank 120 Sodium Hydroxide Storage Tank c 408 Reactor Coolant System 9,146 I

Minimum Containment Sump Volume = 59,465 ft 3 4

NOTES: (a) Refueling Water Storage Tank at minimum operating water level at start of drawdown of the tank. The tank drawdown will

, be terminated at LO-LO level upon the automatic initiation of recirculation via the RHR system.

i (b) The three (3) Safety Injection Accumulators have individual capacities of 1000 ft3 each.

T 1

(c) The minimum usable volume of liquid in the Sodium Hydroxide Storage Tank.

9 f

-t 9 f % -. i+ . - , ' , - %,= -.n-.-, -w-y

IA31E 12A. 3-2 V. C. Su==er Shielding Source Ter=s (T=0)

Liquid (1) Gaseous (') Containment Containment Source Source Susp Reactor Coolant Airborne 'a'as t e Gas Activity Activity Concentration Concentratica Concentration Concentration Isotoee (ci) (C1) (uci/cc) (uC1/ce) (uci/ce) (uci/ce) 3r-Sa 7.5

• 6 3.75 + 6 a.46 + 3 2.90 - a 7.20 + 1 6.67 + 3 Kr-87 3.70 - 7 3.70 - 7 2.20 - 4 1.43 + 5 7.10 + 2 1.30 - 6 Te-133 a.10 - 5 -

2.62 - 2 1.70 - 3 - -

C4-134 1.70 - 5 -

1.01 - 2 6.56 - 2 - -

C4-136 a.70 - 4 -

2.30 - 1 1.31 - 2 - -

C4-137 7.03 - 1 -

4.13 - 1 2.71 - 2 - - 34-139 1.50 - 6 -

S.93 - 2 5.79 - 3 - -

3r-83 3.20 - 6 1.60 - 6 1.90 - 3 1.2a - a 3.07 - 1 2.35 - 3 Kr-33m 6.30 - 6 6.30 - 6 3.75 - 3 2.43 - i 1.21 - 2 2.09 - 5 Kr-65m 1.90 - 7 1.90 - 7 1.13 - 4 7.34 - a 3.65 - 2 5.06 - 5 Kr-65 7.a2 - 5 7.42 - 5 a.a2 - 2 2.36 - 3 1.42 - 1 1.15 - a Kr-68 3.50 - 7 5.50 - 7 3.27 - a 2.12 - 5 1.06 - 3 1.63 - 6 15 -38 5.41 - 5 -

3.22 - 2 2.09 - 3 - -

15-69 6.30 - 5 -

4.05 - 2 2.63 - 3 - -

Sr-89 7.40 - 5 -

4.10 - 2 2.36 - 3 - -

Sr-90 5.00 - 4 -

2.98 - 1 1.93 - 2 - -

T-90 5.00 - 1 -

2.98 - 1 1.93 - 2 - -

3r-92 9.10 - 5 -

5.42 - 2 3.51 - 3 - -

Y-92 1.00 - 6 -

5.95 - 2 3.36 - 3 ,

Sr-93 1.10 ' 6 -

6.55 - 2 a.25 - 3 - -

Y-93 1.10 - 6 -

6.55 - 2 '.. 25 + 3 - -

Mc-99 1.50 - 6 -

3.93 - 2 5.79 - 3 - -

2c-99 1.30 - 6 -

7.74 - 2 5.00 - 3 - -

lu-103 1.10 - 6 -

6.55 - 2 4.25 - 3 - -

Ah-103m 1.'O - 6 -

6.55 - 2 4.25 - 3 - -

la-106 '.10 - 3 -

2.a4 - 2 1.58 - 3 - -

Rh-106 a.10 - 5 -

2.44 - 2 1.33 - 3 - -

Te-132 1.lc - 6 -

6.55 - 2 a.25 - 3 - -

2-132 5.90 - 7 2.95 - 7 3.51 - 4 2.23 - 5 5.66 - 2 5.24 - a Te-13a 1.60 - 6 -

9.52 - 2 6.13 - 3 - -

2-134 3.90 - 7 a.45 - 7 5.30 - a 3.44 - 5 5.54 - 2 7.58 -

12A-7

TA3L 11A.3- 2 (Cca:inued)

Liquid ( Gaseous ( } Containmen: Contain=en Source Source Sc=p Reactor Coolan: Airbo rne 'aaste Gas Ac:ivi:y A :1vity Concent:stion Concentration Concentration Concen::ation 2so:cee (ei) (ci) (uci/ce) (uci/ce) _ uci/ce)

( (uci/ec)

Xe-138 1.50 - 3 1.50 - 3 S.93 - a 5.79 - 5 2.38 - 3 6.64 - 6 Cs-138 1.50 - 6 -

S.93 + 2 5.79 - 3 - -

3a-140 1. ' O - 6 -

S.33 - 2 5.41 - 3 - -

La-la0 1.50 - 6 -

S.93 + 2 5.79 - 3 - -

Ce-la3 1.20 - 6 -

7.la - 2 4.63 - 3 - -

?:-143 1.20 - 6 -

7.14 - 2 a.63 - 3 - -

Ce-laa 9.20 - 5 -

5.43 - 2 3.35 - 3 - -

? -laa 1.20 - 6 -

7.la - 2 ' 63 - 3 5 -91 9.20 - 3 -

5.a3 - 2 3.35 - 3 - -

T-9t= - - - . - -

?-91 9.60 - 5 -

5.71 - 2 3.71 - 3 - -

2:-95 1.30 - 6 -

7.74 - 2 5.02 - 3 - -

Nb-95= - - - - - -

Nb-95 1.30 - 6 -

7.7a - 2 3.02 - 3 - -

2 -97 1.30 - 6 -

7.74 - 2 5.02 - 3 - -

Nb-9 7: - - - - - -

Nb-97 1.20 - 6 -

7.74 - 2 3.02 - 3 - -

Ru-105 3.0-5 -

5.12 - 2 3.32 - 3 - -

Rh-105c $ S.60 - 5 i -

5.12 - 2 3.32 - 3 - -

Ah-105 5.30 - 3 -

3.27 - 2 2.12 - 3 - -

.a

. _ .3 . i ..-aa .o _

a..o. .

. . . < S- a ., _ _

.0_- , ,c _ - , > . , ,

._.t., , .. . ..- _ 4 ...=3 2

, 3.o.; . ,. 3. _ -

Xe-131= 6.31 - 5 6.31 - 5 3.38 - 2 2.51 - 3 1.25 - 1 1.13 - 4 2-133 '.30 - 7 3.90 - 7 .6a - a 3.01 - 3 7.'9 - 2 6.91 - 4 Ie-133= 3.30 - 6 3.30 - 6 2.26 - 3 1.47 - 4 7.29 - 1 7.2 -a Xe-133 1.50 - 3 '.30 - 3 3.93 - 4 3.79 - 5 2.38 - 3 2.72 - 6 i

!  :-135 6.90 - 7 3.a5 - 7 a.11 - a 2.66 - 5 6.62 - 2 6.09 - a Ie-135= a.20 - 7 a.20 - 7 2.50 - a 1.62 - 5 S.C6 - 2 1.77 - 6 i.

X4-135 2.90 - 7 2.90 - 7 1.73 - 4 1.12 - 5 5.37 - 2 7.13 - 5 t

4 3a-lal 1.30 - 6 -

7.74 - 2 5.02 - 3 - -

La-141 1.30 - 6 -

7.7a - 2 3.02 - 3 - -

<e-t+&

. .-0 - e. -

s.33 c . ., - 3 - -

i 12A-3 I

i

TA3LI 12A 3-2 (Con:inusd)

NOTIS:

(1) Based On 1000 noble gas core inven:ary, 50 halogen c:re inven: cry, and i: ef all c:hers ec:e invent:ry.

(2) Based en ICO: nobit gas core inven:er7 and 25 halogen core inventory.

12A-9

_ _ .