ML20127L710

Forwards Response to Request for Addl Info,Including Draft SER Open Item 51, Toxic Gas Evaluation of Chemicals. Info Will Appear in Amend 16 to FSAR

ML20127L710
Person / Time
Site:	Vogtle
Issue date:	04/03/1985
From:	Bailey J GEORGIA POWER CO.
To:	Adensam E Office of Nuclear Reactor Regulation
References
GN-574, NUDOCS 8506280100
Download: ML20127L710 (42)

Text

- - - - - . - . . . _ . ..

I h Georgi 1 Power Company Rout 12 Box 299A Waynesboro, Georgia 30830 Telephone 404 554 9961 g

404 724 8114 Southern Company Services, Inc.

m Post Office Box 2625 Birmingham, Alabama 35202 Telephone 205 8704011 Vogtle Proj.ect April 3, 1985 Director of Nuclear Reactor Regulation File: X3BC35 Attention: Ms. Elinor G. Adensam, Chief Log: GN-574 Licensing Branch #4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555 s

NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 V0GTLE ELECTRIC GENERATING PIANT - UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION - DSER OPEN ITEMS

Dear Mr. Denton:

3 Your staff, as part of its VEGP review process, has requested additional information. Attached is an inder of_the enclosures which address your staff's requests. As noted in the remarks column of the attachment, this information will appear in Amendment 16 of the VEGP FSAR.

If your staff requires any additional information, please do not hesitate to contact me.

Si cerely, f.

J. A. Bailey Project Licensing Manager JAB /sm Enclosure rc: D. O. Foster R. A. Thomas G. F. Trowbridge, Esquire J. E. Joiner, Esquire C. A. Stangler L. Fowler M. A. Miller L. T. Gucwa {y i

G. Bockhold, Jr. l 0138m 8506280100 850403 PDR E

ADOCK 05000424 PDR 0 60

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Af O

s: Attachment Open Item Enclosure Remarks 32 A Loose parts monitoring.

VEGP commitment to con-formance to all positions of R.G. 1.133. This revision will appear in Amendment 16.

51 B Toxic gas evaluation.

Response provided to NRC letter dated 3/11/85 with FSAR revisions to incorporate offsite evaluations. These revisions will appear in Amendment 16.

58 C Isolators used in BOP design. Supplemental information requeted by the PSB during the March 22, 1985 NRC meeting.

103 D SGTR. Response to the specific concerns expressed

, in Q440.129. GPC will address the overall concerns of SGTR as part of the Westinghouse Owners Group.

115 E Tornado missile protection for cooling tower fans.

Supplemental informa-tion requested by the ASB during the March 6, 1985 teleconference.

Confirmatory Item 1 39 F Type C testing of penetrations 30-39 and 56-60. Revisions to T6.2.4-1, para. 6.2.6, Q480.33, and Q480.24.

These revisions will appear in Amendment 16.

! 0138m

f . _ . _ _ _ .1F enchsure A s oI -3 2.

VEGP-FSAR-1 1.9.132 REGULATORY GUIDE 1.132, REVISION 1, MARCH 1979, SITE INVESTIGATIONS FOR FOUNDATIONS OF NUCLEAR POWER PLANTS C 1.9.132.1 Regulatory Guide 1.132 Position Paragraph C of the guide addresses site investigations for foundations.

1.9.132.2 VEGP Position VEGP site investigation conforms with the requirements of this regulatory guide. Refer to section 2.5 for discussion on this subject.

13 1.9.133 REGULATORY GUIDE 1.133, REVISION 1, MAY 1981, LOOSE-PART DETECTION PROGRAM FOR THE PRIMARY SYSTEM OF l0 g

LIGHT-WATER-COOLED REACTORS

{) 1.9.133.1 Regulatory Guide 1.133 Position This guide describes a method asceptable to the NRC for implementing requirements with respect to detecting a .

potentially safety-related loose part in light-water-cooled reactors during normal operation.

1.9.133.2 VEGP Position Coderm, yRefer to subsection 4.4.6.4 for a discussion of the digital metal impact monitoring system (DMIMS) which is the VEGP loose part monitoring system. ? CT ;;nf;;n; t; n;;;lat; , C_ ;_ 0 1.133, ..th felle i.., clm..fication; t; rie.. sic.. C.C. "p e . .

{j rc::ipt :f 2r. 212:n, "ECP fill in"::ti; t the 21:rr tr : -fir-13 15

.f ; 1;;;; :;rt :::i r t e . "n er;ine ri ; r"21r-*i^- cf ^ nfir--e i____ part; ill b; p;rf;ra;d t; d;tcrain; ah;thcr ; ;;p;rtshi;

nditi;n h;; ;;;;.. l me deem Led u 10 CIA 50.72 and 10 CER 0.7^. i;GZ ouasi Zvalvw Luc & wqu t i au.a u L a va 10 CER 30.72 and 10 CCR 50.73 fe. y . J. . ., y . m.. et notification and f;110-up (j r;;

rtin; cf the renfirr

ti n ^# - 1^^-^ r'**

Amend. 10 9/84 Amend. 13 1/85 1.9-103 Amend. 15 3/85

lf bc,le54Pe.6 i,

k' ADDITIONAL INFORMATION NEEDED TO RESOLVE OPEN ITEM 51 "T0XIC GAS EVALUATION OF CHEMICALS"

1. Identify the important provisions of emergency procedures you will develop for coping with toxic gas release accidents, taking into account Regulatory Positions C.15 an C.13 of Regulatory Guide 1.78.

2. Additional information is needed to clarify an inconsistency between Table 1 of Regulatory Guide 1.95 and FSAR Table 2.2.3-19 relative to control room leaktightness which will be periodically verified by requirements in the Technical Specifications. Table 2.2.3-19 shows an isolated inleakage of 1500 ft3 (sic). The second paragraph of FSAR Section 6.4.4.2, on the other hand, implies that the control room is designed to the isolated inleakage of 169 CFM (0.06 air changes per hour) as called for in Table 1 of Regulatory Guide 1.95.

Provide detailed justification if isolated inleakage assumed in the FSAR is greater than that called for in Table 1 of Regulatory Guide 1.95.

Response

1. The provisions necessary for coping with a toxic gas release, such as chlorine, are addressed in plant procedures. These provisions include measures such as the following:

[ e Action to be taken upon detection of chlorine in the Control Room I or TSC areas, "

e Isolation of the source when found, e Isolation of control room hvac and switching to emergency mode, and e Evacuation of non-essential personnel from the Control Room by the Shift Supervisor and wearing breathing apparatus until chlorine is within aceptable range.

In addition, arrangements with the appropriate Federal, State, and local agencies and other cognizant organization for the prompt notification of the plant in the event of a hazardous chemical accident within five miles of the plant have been made. See FSAR Section 1.9.78 for VEGP conformance to Regulatory Guide 1.78.

2. Regulatory Guide 1.95 requires that control room operators be protected from an accidental chlorine release and that adequate protection is provided such that the Regulatory Guide 1.78 toxicity limit of 15 ppa C1 is n t exceeded in less than two minutes.

2

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Table 1 of Regulatory Guide 1.95 presents six control room types covering the expected range of protections required for calculating the maximum allowable C1 invent ry in any single container as a 2

function of its distance from the control room. In addition, the text of this Guide states: "Other combinations of the significant parameters (in Table 1) are possible, but those listed in the table should provide sufficient guidance in most cases."

While the intent of Table 1 of Regulatory Guide 1.95 is to cover all expected ranges within the six control room types, industry experience on San Onofre 2&3 and ANPP has shown that infiltration rates generally exceed 0.06 air changes per hour.

VEGP was originally designed as a type 1 control room (0.06 air j changes per hour) per Table 1 of Reglatory Guide 1.95. However, based on testing experience and Ie~ sign modifications to other similar plants, the emergency HVAC units were increased in size to 1500 CFM (0.52 air changes per hour) to ensure that a 1/8" positive pressure could be maintained within the control room during the emergency mode.

As a result, a deterministic chlorine analysis (discussed in FSAR paragarph 2.2.3) was performed using the data in FSAR Table 2.2.3-18. The results of this analysis meet the acceptance criteria of Regulatory Guide 1.95 (operator protection from chlorine releases) and Regulatory Guide 1.78 (15 ppa C1 2 toxicity limit not exceeded in less than two minutes) and are therefore acceptable.

As a result of these design evolutions, the following FSAR paragraphs will be revised: 1.9.95.2, 2.2.3.1.4.1.1 and 6.4.4.2.

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l VEGP-FSAR-1 e

1.9.95.2 VEGP Position Conform nrr:pt f:r chierin: :t:r_;: dicterc5F~as discussed in g, t paragraph 6.4.4.2. Tr.; " CT ::ntrol r;;;. .. T,y; ! ?- Refer to B.

subsections 2.2.3 and 6.4.2. l I

1.9.96 REGULATORY GUIDE 1.96, REVISION 1, JUNE 1976, DESIGN OF MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEMS FOR BOILING WATER REACTOR NUCLEAR POWER PLANTS Not applicable to VEGP. -

1.C.97 REGULATORY GUIDE 1.97, REVISION 2, DECEMBER 1980, INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS DURING AND FOLLOWING AN ACCIDENT 1.9.97.1 Regulatory Guide 1.97 Position -

This guide describes an acceptable method for complying with

' NRC regulations to provide instrumentation to monitor plant variables and systems during and following an accident in a light-water-cooled nuclear power plant.

1.9.97.2 VEGP Position VEGP conformance is as described in section 7.5.

1.9.98 REGULATORY GUIDE 1.98, MARCH 1976, ASSUMPTIONS USED l FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A RADIOACTIVE OFFGAS SYSTEM FAILURE IN A BOILING WATER REACTOR t

i Not applicable to VEGP.

l 1.9.99 REGULATORY GUIDE 1.99, REVISION 1, APRIL 1977, EFFECTS OF RESIDUAL ELEMENTS ON PREDICTED RADIATION DAMAGE TO REACTOR VESSEL MATERIALS i

1.9.99.1 Regulatory Guide 1.99 Position This guide describes general procedures acceptable to the NRC I for predicting the effects of the residual elements copper and 1.9- 75 Amend. 9 8/84

_ _ . . f k

s VEGP-FSAR-2 f.

where:

p = probability of an accident in the 10-year period within the 30-mile segment.

n = total number of shipments during 10 years = 1200.

i = number of accidents = 0.

The probability density, Pa, of an accident per year per mile t is assumed uniform and can be calculated according to the following formula:

p = JL a Lt where:

L = 30 miles.

t = 10 years. ,

Assuming i = 0.05, 0.5, and 0.95, the lower, median, and upper limits of the accident rate for river transportation are

( given in table 2.2.3-5. The results are close to the median for the nationwide accident rate of 18 x 10-s per year per mile, given in WASH-1238.828 2.2.3.1.1.4 Potential Hazard from Substances Transported Within 5 Miles of the Plant. This paragraph discusses physical and chemical properties of all substances that are transported within a 5-mile radius of the plant site and identifies those substances that can create potential hazards for the VEGP.

Table 2.2.3-6 lists these substances 2nd th: p;tentirl h;;;;dsr-

.; ch;un in tabla 2.2.0 O g ;t:ntir.1 L.;e.-d. ... eu..d

,, e..1 3 A i

(

The potentially hazardous substances are considered in the following paragraphs.cte-ts:

l 2.2.3.1.1.5 Frequency of Accidents with Hazardous Chemicals.

( A. State Highway 23 The frequency of highway accidents with spills can be calculated by the formula:

f sca = XsNcs P, ,

2.2.3-3

. ~ - -

k l

VEGP-FSAR-2

{

According to references 3 and 7, the estimated range for x c is from 2.61 x 10-8 to 6.49 x 10-8 Conservatively, the fraction of trucks with hazardous chemicals becomes:

('

10-8 <xe < 10-8 i; Using this assessment, we find that x s is as shown in table 2.2.3-7.

i According to data from the Bureau of Motor Carrier Safety of the U.S. Department of Transportation, the fraction of spills is 0.064. In the case of gasoline, this fraction varies from 0.02 to 0.3.<ts:

Assuming a legnormal distribution for gasoline, the median value is 0.077, which is close to 0.067. It is assumed that the spill fraction distribution for gasoline is representative of other chemical rel. eases.

Hence, it is assumed for all releases that xs=

"~'

0.067 as a median with an error factor of 4. Trhl: -

3 0.2.2 0 .he.. llc . .uli. vf LL=== w lcul_tirnr. -

5 B. Railroad The Information Sciences Corporation'1** provided data about rail accidents resulting in a spill in the State of Georgia. Data frem the Georgia Public Commission give the number of rail accidents in the State of Georgia. This information is contained in table 2.2.3-9. For calculation of the probability of

.a spill resulting from an accident, it was assumed that at least one car in the train carries toxic chemicals. Table 2.2.3-9 shows these data and results of this calculation.

From this data the lognormal distribution is:

( Median Upper Lower 0.0171 0.021 0.026 The spill rate as provided in an Association of American Railroad Report'18' is 0.152 x 10-8 loss of

( loading per tank car mile. This spill rLte is based on data from 1965 through 1970 (6 years) when a total of 49 loss-of-loading accidents were observed.

(

2.2.3-5

a-t VEGP-FSAR-2 During the years from 1968 to 1978, the nationwide p rail accident rate was 10.9 x 10-s. From this data, the probability of spill, given an accident, is:

-6 0.152 x 10 10.9 x 10

-6

= 0.014 ff /

Table 2.2.3-10 shows the frequency of accidents o railroads with toxic chemicals.

l C. Savannah River Because the speed of barges is much less than the speed of trains and because there is no data ind' . ting accidents with spills on the river, it is c o a.. .vatively assumed that the conditional probability of a spill on the river is the same as on l a railroad (0.021). Table 2.2.3-11 shows the assumed -

frequency of accidents involving toxic chemicals on the Savannah River.

2.2.3.1.2 Explosions There are two classifications af hazardous materials (detailed in table 2.2.3-12) being shipped on routes past the VEGP site which can pose an explosion hazard: flammable liquids and flammable compressed gases.

2.2.3.1.2.1 Allowable Distance and Actual Transported Distance. Regulatory Guide 1.91 describes a method for determining the distance between transportation routes and structures beyond which any explosion that might occur on these transportation routes is not likely to have an adverse effect on plant operation or to prevent a safe shutdown. The distance is based on a level of peak positive incident overpressure below which no significant damage would be expected. This pressure is conservatively chosen at 1 psi. A safe distance can be defined by the relationship:

R 2 KW'1/3' where:

R = distance in feet from an exploding charge of W pounds of TNT.

K = 45, when R is in feet and W is in pounds.

2.2.3-6

4 . -

. _ _. . _ _ l f 6

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^

, 'N . VEGP-ESAR-2 2.2.3.1.4 ' Toxic ' Chemicals t

7.2.3 l'.4.l_' Release of Toxic Chemicals Due to Transportation Accident. There are two classifications of toxic chemical Esterials thatican pose as hazardous chemicals - toxic compressed gases and toxic liquids - being shipped on routes past the VEGP. These chemicals are listed in table 2.2.3-6. ,

Addition 61 chemicals transported to and from the Savannah River ,

Plant'are listed in table 222.2-5.

Three typ Jof releases are ' defined for the toxic hazard eva'luation mathodology:

A. . Gor high vapor pressure liquids such as ammonia and liquid CO,,the amount of material flashed as a result of'depressurization is estimated. This flash release comprises the " puff" portion of the release for sued liquids.

B. The rate of evaporation of high pressure liquids after flashing is also estimated. This release is ,

6 sed asDthe continuous component of the high vapor pressure liquid source.

~

- C. For.nprmal boiling. point This liquidsjthe rate of evaporative release vapothtion is estimated.

comprises the entire source for such substances.

Toxic chemicals trans' ported near the VEGP site are analyzed using the same methodology.as for onsite toxic hazards. This me thodo l o g y i s de s c r i be dgr((ENot4em. 2 . 2 . 3 .1. 4 . 3 . The analys'is assumptions of[m ~es t ict 2.2.3.1.4.3.1 also apply.

For all postulated releases except chlorine, gasoline, ammonia, and nitric acid, the average concentration over an 8-hour period does not exceed the long term toxicity limit value. For chlorine, gasoline,.6nd ammonia, it is demonstrated that there is more than 2 minuten between detection and reaching the snort term toxicity limitGralue as defined in Regulatory Guide 1.78.

Thus, operators have sufficient' time to don protective breathing apparatus prior to exposure to. incapacitating

~

cor,centrations of these toxie gases. Nitric acid slightly exceeds the long term toxicity limit, however, footnote j of table 2.2.3-13 provides justification for acceptability based on the incapacitat g criterna.

lort i The results of this analysis are presented in table 2.2.3-13.

It shows that the control room will remain ha'oitable for all -

release scenarios.and that, only for chlorine, ammonia and gasol*tne.is operator action rerluired.

1 2167t ,

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VEGP-FSAR-2 2.2.3.1.4.2 Potential Hazard from Maior Depots or Storage Areas. The only major depots or storage areas within 5 miles of VEGP are those at the Savannah River Plant and the combustion turbine plant. The chemicals stored at the Savannah River Plant are provided in table 2.2.2-5, and the oils and solvents stored at the combustion turbine plant are provided in table 2.2.3-17. The Savannah River Plant borders the Savannah River for approximately 17 miles opposite the VEGP site. (See subsection 2.2.2.) The combustion turbine plant is located approximately 5000 ft from the VEGP power block.

The chemicals stored at the combustion turbine plant with the exception of the fuel oil No. 2 are stored in small quantities. Due to the fact that these oils and solvents are relatively non-volatile and non-toxic, there is no potential hazard to the control room habitability from these substances.

Fuel oil No. 2 tanks for the combustion turbine plant are located east southeast of the power block, approximately 1350 meters distant. There are three tanks with a capacity of 3,000,000 gal. each. These tanks are surrounded by a dike .

which would prevent the fuel oil from spreading into a large spill area.

The toxic chemical sources at the Savannah River Plant and the combustion turbine plant (table 2.2.3-14) are analyzed using the same methodology as for onsite toxic hazards. This meth f/IMNYgjologyandtheassumptionsusedaredescribedin a Piaae-2.2.3.1.4.3 and 2.2.3.1.4.3.1. The average concentration over an 8-hour period does not exceed the long term toxicity limit except for ammonia and chlorine. For these two chemicals, it is calculated that there is more than 2 minutes from detection to the time that the short term toxicity l limit value is reached. Operators, therefore, have sufficient time to don protective breathing apparatus prior to being exposed to incapacitating concentrations of toxic gases.

The results room The control of thiswillanalysis remainare presented habitable for allinchemicalsye table 2.2.3-}5.g,p cc r- ,gJfor ie r"f f-icient t ime t

, c lorine and ammonia (_dE dict. cacc there j operator /Vto,t ke emergency action.

Rym 2.2.3.1.4.3 Potential Hazard from Onsite Storage Tanks. The storage facilities on the VEGP site are listed in table 2.2.3-18 and are shown in figure 6.4.2-2. The table lists the chemicals, quantities, and their distances from storage to the air intake of the control room.

Several of the chemicals listed in table 2.2.3-18 are excluded Those from further consideration due to their properties.

chemicals excluded are:

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4 VEGP-FSAR-2 A. Lube oil because it is relatively non-toxic and essentially nonvolatile in the absence of aerosolization.'

B. Oxygen because it does not present a potential hazard for control room habitability.

C. Catalyst (benzoyl peroxide) because the melting point ,

is 108'C; so under ambient conditions it is in the solid state. Also because

it is stored in quantities

< less than 100 pounds.'

D. Dispersant (NALCO 7319) because vapor of this liquid is nontoxic.'**'

E. Electrohydraulic control fluid (phosphate esters) because no harmful vapors evolve from this chemical under normal operating temperatures.'

F. Liquids stored below ground level because significant -

spills cannot occur.

G. Seal oil because it is relatively nonvolatile and-nontoxic.'

H. Promotor (dimethyl-p-toludine) becaus'a it is relatively nonvolatile and will only emit NO, vapors when heated to decomposition.'

f I. Hydrogen as a compressed gas because it is stored in quantities less than 100 lb.

J. Sodium hydroxide since it only poses a threat if it is inhaled in the form of dust or mist in the immediate vicinity of the spill. The control room is sufficiently distant to preclude the inhalation of dust or mist.'

The chemicals that are analyzed are divided into three categories: (1) compressed gases; (2) liquified compressed gases and liquids with boiling points below the ambient temperature (low boiling point liquids); and (3) liquids with boiling points above the ambient temperature (normal boiling i point liquids). These chemicals can either emit toxic vapors f or can be'asphyxiants. Liquids can be both pure substances or l

aqueous solutions. Compressed gases include nitrogen and sulfur dioxide. Liquified compressed gases and low boiling point' liquids include Halon 1301, carbon dioxide, chlorine, hydrogen, and nitrogen. Normal boiling point liquids include ammonia (29)"

3 2167t

JF 5

VEGP-FSAR-2 llh Fires  !

2.2.3.1.5 t

In the vicinity of VEGP, the following potential fire hazards exist:

A. Fire due to a transportation accident.

B. Fire due to an oil or gas pipeline rupture accident.

C. Forest fire.

D. Fire due to an accident at industrial storage facilities.

2.2.3.1.5.1 Fire Due to a Transportation Accident. Due to the distance from the plant to State Highway 23 (26,500 ft) and l1':

the small quantities of flammable materials transported along 5 this route, there is no potential hazard to control room 3 habitability from an accident with a fire along this transportation route. .

The distance from the plant to the Seaboard Coast Line Railroad g

is 23,900 ft. As c ' ~>- 1" *'kir 2.2.2 12, sulfosiu-ecid *"i*.

1rr.-u._ req"4*- cenrid-**+4mn- =4n - their . Lual m... ox;: dr '--

, th; ";;ulatery Cuide 1.70 all;w ble mass. En1fuv4- :cid ir n:4 ~

! -fl : :ble, +h-refer , n:t : fire h222rdflLThe flammability limit ,

of ammonia is in the range of'15.5 to 27.0 percent volume concentration.'18' These high limits cannot be attained in open spills.

l The distance from the plant to the Savannah River is more than g 105Q - 700 fisY Major cargo transported along the Savannah River

^

I i includes gasoline and No. 2 fuel oil.

The maximum quantity of petroleum fuels transported on the Savannah River is 2.2 x 10' lb. The flammability limits of l1!

t gasoline are 1.3 to 6 percent volume concentration.'2** The

, flammability limits for No. 2 fuel oil are 1 to 7 percent volume /

concentration.'888

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d mNula L5 T k Onl} [l(w m A k M 4 b V M I ~/f d e f[(V b C!

M0] i bl( 5 Vd d f. 4.ha.k e p ire.s fwe h r c.o ns ;ct., ken, 1

Amend. 3 1/84 2.2.3-17 Amend. 15 3/85

T 4

VEGP-FSAR-2 37.- Slide, D. H., " Meteorology and Atomic Energy 1968,"

Prepared for the U.S. Atomic Energy Commission, July 1968.

38. Hazer, Dr. Kathy, Toxologist, American Petroleum Institute, Telecenference, August 17, 1980.

39. EHC Fluid Specifications and Maintenance Pamphlet, GEK-46357A.

40. Nalco Chemical Company, NALCO 7319 Material Safety Data Sheet, 1980.

41. Sax. N. Irving, Dangerous Properties of Industrial Materials, 6th Edition, Von Nostrand Publishing Co. New York, 1984.

42. Back, K. C. and Thomas, A.'A., Aerospace Problems in Pharmacology and Toxicology, pp. 395-411, 1970.

43. American Conference of Governmental Industrial Hygienists, "TLV's Threshold Limit Valves for Chemical Substances and .

Physical Agents in the Workroom Environment with Biological Exposure Indices with Intended Changes for 1984-85," 1984.

44. Proctor, N. H., and J. P. Hughes, Chemical Hazards of the Workplace, J. B. Lippincott Company, Philadelphia, 1978.

45. Sittig, Marshall,. Handbook of Toxic and Hazardous Chemicals, Noyes Publication, Park Ridge, New Jersey, 1981.

46. American Council of Government Industrial Hygienists, l

" Documentation of the TLV's for Substances in Workroom Air" Third Edition, Cincinatti, Ohio, 1976.

(

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. TABLE 2.2.3-6 _

CHEMICALS TRANSPORTED ALONG MAJOR TRANSPORTATION ROUTES WITHIN 5 MILES OF VEGP - QUANTITIES AND DISTANCES Minimum Maximus Distance Mode of Shipment From Control Chemical Transport Physical Conditions Size Room Air Intake Chlorine Truck Liquefied, compressed 1 ton cylinders 7600m gas; ambient conditions Truck Liquefied, compressed 6 tons 7600m Anhydrous ammonia gas; 28*F and 250 pst Nitrogen Truck Liquefied, compressed 6500 gallons 7600m (11guld) gas Phosphoric Truck Liquid, pure state, 200 lb stain- 7600m acid ambient conditions less steel drums -

Nitric acid Truck Liquid; pure state 5000 gallons 7600m ambient conditions No. 2 Truck Liquid, ambient 6000 gallons 7600m diesel fuel conditions ~

oil Anhydro s Rail Liquefied compressed 26 tons 7250m ammonia sj F and Sulfuric Rail Liquid, pure state 13,400 gallons 7250m acid at ambient conditions Rail Liquefied compressed 20 tons 7250m Carbon dioxide gasj

'C & 300 pst Hellum Rail Liquefte pressed 200,000 ft' 7250m gasjambient temp, 2200 pst Nitrogen Rail Liquefied compressed 14,000 gal 7250m (11 quid) gas No. 2 Barge Liquid; ambient 2,200,000 gal 1050m diesel conditions fuel oil Gasoline Barge Liquid; ambient 2,200,000 gal 1050m conditions 2168t

_ ..-~ ,.

TABLE 2.2.3-8 TRCO"E"CY OF ?.CCIDENT "IT" "n?nof0"C C"E"IChrC

-Gii 3 TATE "!gutJav 73 -

.- Accident Rate Frequency of Accident per Mile per Shipment per Nile per Yea r Probabi1e Number or .

Eu_hn a. nge _of_Sgi11 h_ipmen u (gw_e r fie_d_i_a r) Upper IJwer- Hedian Upper

Hydrogen sulfide 0.067 7 1.06 v 10' 1.73 x 10 1.03 x Iri' fa.97 y 10-# 8.840 x 10*# 1. f42 x 10**

~#

Phosphoric acid 0.n61 1 1. 0bqu 1.13 x 104 1.03 x 10-6 p,33 , 99-7 3.60 / 10 6.09 x 10'#

Ammonia 0.0(7 2 1.06x'1[e' .1.73 x 10~" 3.03 x 104 1. f42 y 10 ~#  ?. f40 x 10'# I.06 x 10*#

4 Fuel oil 0.06 01~ 1.06 10 4

1.73 x tr' 1.03 10 6.89 x iri' i.itx 10-5 1.9 7 x 10~5 4 4 Nitric acid r.T137 12 1.06 x 10 1.73 x 10** 3.03 x 10~' 8.52 x 10'# 1.14 r4 x 10 2. 84 ts x trre hinrinn 0.067 I? 1.06 x 10 1.73 x 10~' 3.03 x 10~' 6.52 s 10'# 1. fa ts y 10 '

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TABLE 2.2.3-10 FEEQFTEhfCV rw ArrTrWMT IJ T 'ru un' inn g rig c;;;,u;;;; %

0;; RAILRGADG IN OCO"OI A--

}

i .

Accident Rate F rerstmncy of Accident per Mile p r Shipme.nt_ pe r Mi le pe r Yea r .

A

! Probanesii, "-h a r or

,n t, , Sp i.1 1 . Sh i pmen t.-I'~ t_ owe r - Hed[.mn Upper t. owe r Hed,l a r! yppE l j Stehstancy Stelfair ic acid 0.021 1 5.f 0 x 1ri' 4 6.96 e 10 8.9/> r 10 _1,84f t, y 10'# 88.18 x 10-# 5 . 6 84 x 10'I l

8.96 w ifr6 3, ygg y in4 p,39 y g n-S p,gp y in 4 Ammonia 0.D?l _ _ 150 5 . 8 488 < I f t 6.9fs x ItF' 6.96 x 10-' 8.96 y 10 2.55 y 10 4 1.?9 / 10 4 14.23 y 10 4 Carbon 6'~f 21

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1 I ml x.t '

Th4 kW has Leer, cle/ekd. .

w i

'b L

Rep 4ee 4sk j. 7.3.y TABLE 2.2.3-13 TRANSPORTATION SOURCES - T0XIC GAS RELEASE INFORMATION (Sheet 1 of 2)

Control Room Consentrat1on Toxtetty 8-hour Average'88 Maximun' 8 8 8 8 8 Fraction Vapor'" Odor "8 2 Minutes Limit Concentration Release'C' Release Rat Flashed to Pressure Detection AfterDetytton ,

(ppm) (ppm) Type (gm/s) Vapor (sus Hg) { ppm) (ppm) * '

Chestcal TRUCK Anhydrous Assnonia 500 N/A L 4.996 x 10 3 0.18 N/A 50 69 Nitrogen (Liquid) 143.000 0.0 L N/A 1.0 N/A N/A N/A Phospheric Acid 0.25 3.8 x 10-3 N 5.0 x 10~3 N/A 0.0285 N/A N/A Nitric Acid 2.0 2.85 0' N 1.601 x 10-2 N/A 10 N/A N/A No. 2 Fuel Oil 300 "'

8 O.14 N 1.41 x 10' N/A 0.408 N/A N/A Chlorine 15 888 N/A L 3.187 x 10 3 0.18 N/A 2.0 2.9 f

un Anhydrous Assnonia 500 N/A L l.592 x 10* 0.18 N/A 50 112 Carbon Dioxide 5000 41 L 1.187 x 10 3 0.17 N/A N/A N/A Heltum 143,000 35 G N/A 1.0 N/A N/A N/A Nitrogen (Liquid) 143,000 0.0 L N/A 1.0 N/A N/A N/A Sulfuric Acid 0.25 0.0037 N , 2.6 x 10*3 N/A 0.005 N/A N/A BARGE No. 2 Fuel Oil 300 30 N 1.62 x 10 2 N/A 0.408 N/A N/A I Gasoline 500 N/A N 2.585 x 10 4 N /A 403 0.09 262 l

2169t L

• e TABLE 2.2.3-13 TRANSPORTATION SOURCES - TONIC GAS RELEASE INF0aMATION (Sheet 2 of 2) l A. The two-minute textetty limit is presented for chlortne, ammonta and gasoline only. The long term (4-h average continuous exposure)

! textetty limit is presented for all other materials. All values are from reference 43.

l 8. At worst case windspeed.

i j C. N = normal belling point 1tould (boiling point > ambient temperature). Continuous release scenarlo. L = low boiling point Itquid or

llouefied compressed gas (botting point < ambient temperature). Puff release plus continuous release scenario. G = compressed gas

release. Puff release scenario.

4 D. Continuous release rate for normal boiling point Itquids. Bolloff rate for low boilins point.

{,

E. Vapor pressure not used in analysis of low botilng point Itquids or compressed gases.

F. The odor detection limit is only presented for ammonta and gasoline since the analysis considers control room concentrations 2 min after odor detection. Chlorine detection is by instrumenation.

G. From U.S. NRC Regulatory Guide 1.78. June 1974.

H. The value for essoline is used since toxicity Itmits for fuel oil have not been estabitshed.  !

J. The long term textetty limit is based on continuous exposure for'n 40-hour work week and results in eye irritation and teeth eroston. An 8-hour exposure at levels slightly above this ilmit will not incapacitate control room operators. Additionally, the 2-etnute textcity llett valve is never exceeded for nitric acid releases. Furthermore. nitric acid decomposes to nitric oxide and nitrogen dientde in the presence of air.'***. If all the nitric acid is assumed to be converted to (the more textc) nitrogen dioxide the long ters 8-hour textcity llett i

valve is never exceeded. Therefore, even though nitric acid exceeds the long term textetty limit. It will not incapacitate control room operators.

l I

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2169t

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I Rephee Mle

$. 2. 8 -/W wild TABLE 2.2.3-14 CHEMICALS ANALYZED THAT ARE AT OFFSITE i STORAGE FACILITIES - QUANTITIES AND DISTANCES Distance from Maximum Quantity Control Room  !

Chemical Physical Conditions Stored Air Intake Chlorine Liquefied, compressed 34 tons >5000m at Savannah gas; ambient River Plant (SRP) conditions Anhydrous Liquefied, compressed 2 tons >5000m ammonta gas; ambient at SRP conditions Phosphoric acid Liquid; pure state: 460 lb >5000m at SRP ambient conditions Sulfuric acid Liquid; pure state; 270 tons >5000m at SRP ambient conditions No. 2 diesel Liquid; ambient 22,500 gallons >5000m fuel oil conditions at SRP No. 2 diesel Liquid; ambient 3,000,000 gallons 1350m fuel oil at conditions -

combustion turbine plant.

i l

t 2173t f - _ - ___ - -

. a Replace. /abk A23-5 TABLE 2.2.3-15 0FFSITE SOUSCES - TOMIC GAS RELEASE INF0st4 TION i doif Ji Mis /a$pIc.

i Centrol asem Concentration Toxicity 0-hour Averace Maximus'** Fraction vapor'" Odor'" 2 Minutes Liett Concentratton telease'C' selease sat Flashed to pressure Detection After Detectlen (ppm) (pen) Type (en/s) Vapor (en He) (p0m) (ppm)<an Chestcal ,

AT SAVARRAN RIVER PLANT Chlorine 15'88 N/A L 4.065 x It* 8.10 N/A 2.0 0.20 Anhydrous Ammonta 500 N/A L 2.154 x 10 3 0.10 N/A 50 70 phospheric Acid 0.25 6.9 x 10'8 N 3.5 x 10-3 N/A 0.0205 N/A N/A Sulfuric Acid 0.25 0.015 N 3.6 x 10-' N/A 0.005 N/A N/A Diesel Fuel Oil 300'"' O.0 N 4.75 x 10' N/A 0.400 N/A N/A 4 7 co m e s g ., 7 ,3;,,,, p g ,g.

Diesel Fuel 011 300'"' 4.1 N 1.56 x 10' N/A 0.400 N/A N/A dnd A. The two-minute toxicity limit is presented for chlorineg ammenta rd ; ::"x"only. The long ters (0-h average continuous exposure) toxicity limit is presented for all other materials. All values are from reference 43.

B. At worst case windspeed.

C. N = normal botilne point Ilguld (botilne I,otnt > ambient tensberature). Continuous release scenario. L = low belling point Ilgute or ltquefted compressed ses (botilne point < ambient temperature). puff release plus continuous release scenario. G=

compressed gas release. puff release scenarlo.  ;

D. Continuous release rate for normal botilne point Itquids. Rolloff rate for low boiling point.

E. Vapor pressure not used in analysts of low betilne point Itquids or compressed eases,

f. The oder detection llett is only presented for ammonta rd ; :"t[stnce the analysts considers control recta concentrations 2 min after odor detection. Chlorine detection is by instrumenation.

G. from U.S. NaC seculatory Guide 1.70. June 1974.

H. The value for gasoline is used since toutetty lletts for fuel oil have not been established.

I 2174t

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t.

VEGP-FSAR-2 hd TABLE 2.2.3-16 mm. ERGLABILIT!"O 0" "OEEDINO "E ""Ecunr.n LIMTTc: nitr -

TO T7Jd43FORIAIIQh avu Aui.WI Ol4 SAVA1414A KIVER %

Lower Median Upped Substance '

_7 1.04 x 10-8 l Fuel oil 2.96 x-5.56 x 10-8

,s Gasoline 1.60 x 10-' 1.74 x -' 9.58 x 10-e N._ ~~ . . . _ _ ____ . _ _ _ . _ _ _ _ -

This }able has been del'I'N- -

l l

l i .

l.

i l

(

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1~

b VEGP-ESAR-2 f

TABLE 2.2.3-17 OILS AND SOLVENTS STORED AT THE COMBUSTION TURBINE PLANT Quantity Substance (gal)

Fuel oil No. 2 9,000,000 (,3 - 3poo o** j'I j

, 44nks)

Gulferest 32 700 Gulf diesel Motive 485 700 Gulf Harmony 115 200 Gulf HD SAE 10W 100 Gulf Senate 110 100 Shell 20W Oil 100 .

Gulf Senate 3200 50 t

Momar Electraclean 200

(

(

_ ~ - - ..

TABLE _2.2.3-20 gal irboe ON S ITC ddd M ES g ,gnfta % a TOXIC GAS RELEASE INFORMATION t

, , 3 g g ,,Qcr Ah,e,twa Qpi n a LL=

dm m 6.m. ---

ga3 cuu nuocnany 8-h'oYir- ga3 Odorg ,,

Average gc3 Ma x(e mumfraction Vapor 2 .- :

Y f Chemical Toxicity gy Concentration Release Limit typal f ppm.) J ype Release Rate flashed Lqu/s) lo_ Vapo _r Pressure Detection

_[mmyg L jpfel ' b "; I'y , ;-

3 Ammonia (29%) 500'I*' N/A N x 10 N/A 486 50 -P902$f a D.13 fr.?t (e) N/A N/A Carbon dioxide 5000 4039 L 7.193 x 10 I

Chlorine 15 N/A L 3.187 x 10 (e) N/A 4Pt I si 14+e 0.3827 N/A i fuel oil no. 2 +00' 3ed 1.2 N 4.749 x 10' N/A N/A 70,000'#U 999 L N/A 1.0 (e) N/A N/A Halon 1301 1421 bbN

-lb-e*7 x 10 N/A 20 4 .Wy9- f b e h Hydrazine (35%) 30 N/A N g 0.0 s Hydrogen - liquid 143,000 0.02 L 1.178 x 10' 6 (e) N/A N/A A N/A Nitrogen - liquid 143,000 1731 Lf WA I-Mll l6 (e) N/A Nitrogen - gas 143,000 2772 G N/A 1.0 (e) N/A N/A

- Su^;um ;.W uu un taval 1.4 'O N ' ' " *I# ' ' ' " '; n/n --

.~

Sulfuric acid 0.25 0.16 N 3x 10** N/A 0.005 N/A N/A Sulfuric dioxide 5 1.1 C N/A 1.0 (e) N/A N/A

a. The 2-minute toxicity limit is presented for chlorine, ammonia, and hydrazine only. The long term (8-h averago continuous exposure) toxicity limit is presented for all other ma teria ls. All values are from reference 43 unless otherwise noted.

b. At worst case windspeed.

c. N = normal bolling point liquid (bolling point > ambient tempe ra tu re ) . Continuous release scenario. L= low boiling point liquid or liquefied compressed gas (boiling point < ambient temperature). Puff release plus continuous release scenario. C = compressed gas release. Puff release scenario.

i y d. Continuous release rate for normal boiling point liquids. Dollof f rate for low boiling point, h e. Vapor pressure not used in analysis of low boiling point liquids or compressed gases.

the odor detection limit is only presented for ammonia and hydritzine since the analysis considers control room

- f.

un concentrations 2 min af ter odor detection. Chlorine detection is by instrumentation.

W f. from U.S. NRC Regula, tory Guide 1.78. June 1974. *

)

u, h.

q ct So ls w4-T he va l ue fo r Mee l is used since toxicity limits for fuel os i have stot been establ i shed.

1

- I s

VEGP-FSAR-2 TABLE 2.2.3-21 -

PROBABILITY OF EXCE ING THE THRESHOLD LIMI IN CONTROL ROOM ,

DUE TO ACC NT WITH AQUEOUS S TIONS l, l

Substance Upper dian Lower Ammonia 1.28 x 10-8 .05 x 10-8 1.28 x 10-'

Hydrozine . x 10-5 3.2 10-7 2.58 x 10-5 n ;s -lah le has -been de/ded.

9 I

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. ___ ._ -1F

= i VEGP-FSAR-6 6.4.4 DESIGN EVALUATIONS 6.4.4.1 Radiological Protection The effects of potential radiological accidents are analyzed in .

chapter 15. The radiological protection afforded to the operators in the event of an accident is described in subsections 6.4.2, l 12.3.2, 12.3.3, and 12.3.4 and in section 11.5.  !

6.4.4.2 Toxic Gas Protection Control room protection from the effects of toxic gases is in accordance with Regulatory Guide 1.78 as discussed in subsection 2.2.3. The analysis of potential sources for toxic gases is presented in subsection 2.2.3. The analysis of p,nsite and of f site sources for toxic gases is presented in _u___^_ _inne -

2.2.3.1.4.1 through 2.2.3.1.4.3. These sources are analyzed deterministically and it is shown that either the 8-hour toxicity limit is not exceeded in the control room, or that there is at least 2-min between detection and reaching the short term toxicity limit, such that the operators have time to put on -

breathing appara' us. These results are shown in table 2.2.3-20.

As required by Hegulatory Guide 1.95, control room protection is provided against chlorine which could enter the control room. The specific guidelines of Table 1 of Regulatory Guide 1.95 regarding maximum chlorine inventories as a function of control room type and distance tc the control room do not apply since VEGP does not have a control room classified as type I through type VI as described in this regulatory guide. The guide does however note that other combinations of parameters may be acceptable as shown by analysis. Therefore, the intent of Regulatory Guide 1.95 is met as shown by the analysis in subsection 2.2.3.

5 2167t

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• Em lcsum C

• oI sa Response to PSB Question from March 22, 1985 Meeting i The non-Class lE automatic synchronizing system cannot degrade '

diesel generator (DG) operation for the following reasons:

(1) The automatic synchronizing equipment is completely isolated from Class lE circuits by isolating relays and fuses as previously discussed in response to NRC question 420.10.

(2) The automatic synchronizing equipment used on Vogtle are highly reliable solid state devices that have been proven by years of use in the' power industry. Although considered and treated as non-Class 1E devices, they successfully passed seismic shake tests during environmental qualification tests on the generator control cabinet.

(3) Synchronizing with offsite sources using these devices is done only under non-accident conditions during periodic testing of the DG. Testing is done (per technical specification) only on one DG at a time. Under these conditions, failure of the automatic synchronizing -

equipment has no impact on safety of the plant.

' Furthermore, these devices are used only as backups to the manual operator actions required to synchronize and 1 parallel the DG with the offsite source. They are used only in the DG generator breaker control circuit. This breaker is closed onto an,anergized bus during testing by pressing a pushbutton only after synchronized conditions are established by use of a Class lE synchroscope and associated indicating lights (see 1X3D-BA-M10B, D02D).

(4) After a loss-of-offsite power condition is over, either with or without SIS, transfer of the Class lE 4.16 kV '

load back to preferred offsite power sources is performed manually using fully Class lE qualified synchronizing equipment (synchroscope and indicating lights). The circuit breaker providing power to the 4.16 kV bus from offsite sources has no association with the automatic synchronizing equipment. f 2166t I

)G

.g 'I Eia.tes dra D x oI 103 V y . 1

. l VEGP-FSAR-Q Question 440.129 ,

Paragraph 15.6.3.1 of the FSAR states "chargingipump flow -

increases in an attempt to maintain pressurizer level" and "feedwater flow to the affected steam generator is reduced as a result of primary coolant break flow to that unit." Are any control systems used to maintain these levels in the analysis? 3 If so, justify that their operation which you have assumed is '

conservative and modify table 15.0.8-1,to include them. , ..

ge Has credit been taken for the steam generator blowdown liquid [,

monitor or the condenser air ejector radiation monitor? If so, ,-r modify table 15.0.8-1. ,

d. v

Response

" i

.. subgroup of the Westinghouse Owners Group (WOG) has been l15 form o generically address several of the steam gen r tube rupt iconsing issues which have been rai or NTOL plants. Georg war Company is represent the subgroup by Southern Company Se . The WOG s ed WCAP 10698 in December 1984 in response te nerator tube rupture. Two -

v appendixes are scheduled fo . The first is a response to radiological conse es oft stea nerator tube rupture 15.'

(scheduled for A 1985) and the second esponse to SC overfill ( uled for July 1985). Following c tion of the ~

gener rogram, resolution of these concerns will be essed.

No credit was taken in the SGTR analysis for the operation of the pressuriser e level control' system. However, the proper operation of the steam generator i level control system was assumed since this is conservative for the evaluation '

of the offsite radiation doses for an SGTR. It was assumed that the feedwater flow to the affected steam generator is reduced to offset the primary to secondary break flow to maintain level constant in that unit prior to reactor trip. If credit is not taken for the steam generator level control system, .

Y the additional feedwater will result in a higher level in the affected steam generator at reactor trip, which will result in more dilution and a higher attenuation factor for the primary coolant activity that leaks into the steam generator. Thus, the assumption that the steam generator level control. operates to maintain a constant level is conservative for the offsite 'i '

dose evaluation. Table 15.0.8-1 already includes the operation of this ,

v systee.

No credit was taken in the analysis for the steam generator blowdown liquid monitor or the condenser air ejector radiation monitor. .

Amend. 9 8/84 Q440.129-1 Amend. 15 3/85 ,

A

_ 'ss

?\

el VEGP-FSAR-Q .

Question 440.131 Figures 15.6.3-1 and 15.6.3-4 show a differential pressure of about 1000 psi between the primary and faulted steam generator; Figure 15.6.3-11 shows an increasing water volume

'- at 30 min. T-due to the break flowrate as shown in figure 15.6.3-9 at 30 min. Paragraph 15.6.3.2.1 states, however, that leakage flow through the ruptured tube is assumed to be terminated within 30 min of the initiation of the event. Unless these parameters show discontinuous behavior at 30 min, it would appear that the

'" assumption and the figures are in conflict. Please resolve this.

^

If leakage flow is terminated at 30 min, how is it .

accomplished? Any equipment used should be listed in table 15.0.9-1 and qualified.

If the leakage flow is not terminated at 30 min and since the flow through the steam generator safety valve has approached a non-zero asymptote at this time, it would appear that additional radioactive material will be released to the atmosphere. In 1, this event, you will need to reanalyze the radiological consequences.

.v s Response -

^

g

. A ' subgroup of the We'stinghouse' Owners Group (WOG) has been

, formed to generically address several of the steam generator tube rupture licensing issues which have been raised for.NTOL plants. Georgia Power Company is represented in the subgroup by, Southern Company Services. The WOG submitted WCAP 10698 in December 1984 in response to steam generator tube rupture. Two appendixes are scheduled for this WCAP. The first is a response to radiological consequences of a steam generator tube rupture .

(scheduled for April 1985) and the second is a response to SG-overfill (scheduled"for July 1985). Following completio_n of the s-- generic program, resolution of these concerns will be addressed.

1 s- , ,

Y Q440.131-1 Amend. 9 8/84

.Jf

. Q 9'4 0.t3'A VEGP-FSAR-Q D. Provide the noding diagram used in the analysis.

Justify that sufficient noding is provided to predict head bubble formation or loss of natural circulation in loops for which the steam and feedwater flow has been isolated. _

E. Provide the most limiting single active failure. If ,

the most limiting single active failu're is failure of an atmospheric relief valve to close, operator action to close the block valve may be assumed if justified.

v

Response

t31 See the response to question 440.129

[

~

Q440.132-2 Amend. 9 s.

s l

l

- . ---. 1r L o(esuYO b NRC DSER 01-115 (Loss of all Tower Fans)

1. Statement of Problem The NRC has requested we look at what would happen to basin temperatures if we had a tornado coincident with loss-of-offsite j power (LOP), complete loss of all fans in one NSCW tower and loss of the other NSCW train. The analysis has been broken down into several areas:

e How much auxiliary feedwater supply do we have to maintain the plant at hot shutdown, and what can be done using safety-grade components to extend the period at hot shutdown?

e What are the ambient wet bulb temperatures (WBT's) associated with tornadoes?

I e What would be the performance of the NSCW during hot shutdown, cooldown, and cold shutdown following a tornado?

e What effects, if any, would there be on components cooled by NSCW7 II. Auxiliary Feedwater SuDDIY .

Because we postulate the loss-of-offsite power and no accident, we do not have t'o include allowances for reactor coolant pump (RCP) operation or spillage from a break. However, we still provide an allowance for cooldown to a reactor coolant system (RCS) temperature of 350*F. Also, only the safety-grade portion of the condensate storage tanks (CST'sT is taken credit for, as other connections are not missile protected. On this basis we have:

e 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> hot standby capability with one CST operable.

e 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> (2.3 days) hot standby capability with both I CST's operable.

l For long-term hot standby capability, temporary piping could be l

' installed between the NSCW transfer pump in the operable train and the operable CST (s). This would provide unlimited hot standby capability at the cost of having to chemically clean the steam generators before resumption of power operation.

l III. Ambient Wet Bulb Temperatures Associated with Tornadoes

! Based upon a tabulation of reported tornadoes in the vicinity of VEGP combined with ambient wet bulb temperature (WBT) data over

! the 30 years from 1951 through 1980, the following general l

conclusions have been developed:

e The peak tornado months are April and May.

l l

l 1

2143t

- - .1T NBC DSER 01-115

- (Loss of all Tower Fans) e Tornadoes are generally associated with NBT's 10*F above average for the day (s) on which the tornado (es) occur, with a drop of 10*F in WBT one to two days after, e The median NBT for all tornadoes 1951 through 1980, was 65'F, with 75 percent of all tornadoes on days of 70*F WBT or less.

e For NSCW system (ultimate heat sink) analyses, the design WBT's can be taken as follows:

- Use 65-70*F WBT for the day (s) of the tornadoes

- Use 55-60*F WBT for the days following the tornadoes IV. NSCW Performance Followina a Tornado ,

Consistent with Section I, NSCW system performance has been estimated for the case of one train operation during shutdown with loss-of-offsite power (LOP) and assuming all tower fans are lost due to tornado missile damage. The analysis has been divided into three phases: ,,

o Hot shutdown immediately following the tornado until depletion of auxiliary feedwater supply, followed by 5-hour cooldown to RHR cost in at RCS temperature of 350*F.

e Cooldown from RCS tenFtrature of 350*F to cold shutdown (RCS temperature < 200*F) starting 1-3 days after the tornado.

, o Extended cold shutdown operation starting several days

, after the tornado.

A. Hot Shutdown Immediately After Tornado

Since hot shutdown operation (auxiliary feedwater system

! available) occurs immediately following the tornado, an

l. ambient WBT of 65-70*F should be assumed. For one train j

operation with loss of ,offsite f power and loss of all fans in the operable train, the NSCW basin temperature

' will be 97 to 100*F, or only 2 to 5'F above the nominal 4

design valve of 95'F. The potential effects of basin temperature in excess of 95'F are discussed in Section V.

! B. Cooldown from RCS Tennerature 350*F 1

l Assuming that offsite power is not restored or that j continued operation of hot shutdown is not' possible for j any reason, plant cooldown could be initiated as early as 1-3 days after the tornado. Because of this delay, j

i 2

l i 2143t j

I jr i

i NRC DSER 01-115 l

(Loss of all Tower Fans) i the WBT can be assumed to be 10*F lower than during hot

] shutdown, or 55-60*F. However, the heat load during l cooldown is considerably in excess of that during hot i shutdown to limit the effect on peak basin temperature,

! the RHR system can be periodically cycled to limit the i heat dissipated to the basin in any given time period.

I By so doing, the basin temperature can be kept below 1 l

105-110*F during cooldown. However, the time required to achieve cold shutdown (RCS temperature Iso to 200*F) will be several days. However, because a tornado with loss of all four fans is a special case, the time required to achieve cold shutdown is not a consideration.

C. Extended Cold Shutdown Startina Several Days after the Tuna 19.

l Again because we are looking at a time period several

! days after the tornado, an ambient WBT of 55-60*F can be' assumed. However, the NSCW heat loads are greater than during hot shutdown because of the residual decay heat i loads from the fuel in the reactor. It is estimated l that tha basin temperature during cold shutdown will be l

100 to 104*F, or 5 to 9'F higher than the nominal design

! values of 95*F. The RCS temperature during'this period 1 will be 150-160*F. Both the RCS and the basin i . temperatures will decrease slowly as the fuel residual

decay heat load decreases.

i

! V. Potential Effects of NSCW" Basin Tenneratures in Excess of

! 11*f.

i Basin temperatures in excess of 95*F will have some effect on the i various components cooled either directly or indirectly by NSCW.

l However, these effects will be partially offset by the 17% higher l

NSCW flows which occur because the NSCW basin can be presumed full i at the time the tornado occurs. Specific components potentially affected are:

l l e Degraded performance of the containment coolers and ESF chillers will raise the temperature in the respective cooled areas a few degrees but this will be partially i offset by lower ambient temperatures than'used for HVAC design.

l e Increased operating temperatures for the centrifugal l

charging, residual heat removal, component cooling water, and NSCW pumps. Pump failure should not occur, l

but pump and motor life could be affected.

l e components cooled by the ACCW would not be affected, as i they are generally designed to operate at temperatures as high as 120*F.

3 2143t

. _ _ . _ .-- - _ . -. . - . . . if NRC DSER 01-115

. (Loss of all Tower Fans) e The spent fuel pit (cooled by CCW) is not affected, as pool temperature rise margin is 40*F or more.

i e The diesel generators would not be affected, as they are designed for 105*F cooling water.

i N

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4 2143t

_ _. _ . - , If NRC DSER 01-115

- Response

SUMMARY

i e There is sufficient safety grade auxiliary feedwater supply ,

for 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> at hot shutdown with one CST, and 2.3 days with both CST's.

e Unlimited safety grade auxiliary feedwater supply can be made available by using a temporary connection between the operable NSCW transfer pump and the CST's.

e Ambient wet bulb temperature is 65-70*F for days on which tornadoes occur, and 55-60*F for the several days thereafter.

e Basin temperature for hot shotdown immediately following the tornado is 97-100*F (NS 95'F nominal design value).

e Basin temperature during cooldown can be limited to 110*F by cycling the RHR operation.

e Basin temperature during cold shutdown is 100-105'F With RCS temperature of 150-160*F. ,,

o Centrifugal charging, RHR, CCW, and NSCW pumps will see higher cooling water inlet temperatures, but this partially offset by 17 percent higher cooling water flow rates. Pump operability should not be affected.

o Components cooled by compoHent and auxiliary component cooling water systems not affected.

e Diesel generator operation not affected.

e Containment cooler and ESF chiller performance degraded, but the effect partially offset by reduced heat loads and/or lower than design outside temperatures.

g .

1 2152t

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E d esare. F C I- 39 VEGP-FSAR-6 in the test volume by makeup air, nitrogen, or water (if applicable) through a calibrated flowmeter. The flowmeter fluid flowrate is the isolation valve leakage rate.

l Containment isolation valve leakage across the valve seat is determined in the direction cut of the containment, in accor-

)

dance with the requirements of 10 CER 50 Appendix J for Type C testing, and the test fluid (liquid or gas) is the same as that expected during the accident.

Type C testing of the safety injection lines, ::nt:i ._ _.. .

lin::* residual heat removal lines, high head safety injection

)

lines of the chemical and volume control system, RCP seal injection lines, the containment emergency sump lin'es to the residual heat removal and containment spray pumps, and nuclear service cooling. water lines to and from the containment fan coolers is not performed. The justification for this is that these valves are either normally open at the time of a LOCA or are opened at some time after the accident to effect immediate 10 and long term _ co_re cooling. T "'hcr:f:::, th :: line- =*=-

cen;inu uo y u-ter fill:d during m;rgency ces. vuoling y e tenc~,

{- Q:re Livsa wi tI. s f r - . thu sufuclAssy wawcr a6vsage La44h vs favm c pr :nd :: ruch de not pi nide a~

th: centeawaent emergency -

cr:dible path fc. le k.y. ef th: Ocntain-ent tmaphere? T Eurthermore, inservice testing and inspection of these isolation 1 valves and the associated piping system outside the containment is performed periodically under the inservice inspection require-ments of ASME XI as described i rsubsection 3.9.6 and section 6.6. During normal operation, the systems are water filled, and degradation of valves or piping is readily detected. Containment penetrations not drained during Type A testing are identified in table 6.2.6-1.

In the chemical and volume control system, the isolation valves in the charging line are Type C tested using water. This is justified in that these lines are filled with water in the event of a LOCA.

Isolation valves connected to the secondary side of the steam generator, such as main steam isolation valves, main steam relief valves, feedwater valves, blowdown lines, and blowdown conh;n,,ul q,g g y Dnst. systems ars closed systems outside MCu) sys/em which is s. CIAsed syslem irwih er,hinmerd, /

}

destjned efoiremenls, Qad cons /ruc/n/ fo AsmE.221, class .1 sid deism /c 6e/ goy / s(7Mos?Aere and as sac /r Meg s's no/ Cons 6/ule + po/enfia/ conbonmen uij//, & sij/*

leak peM during or A/han.nf a loss-el coo /an/ seciden/the robe lem Componal. SAsu/d 4clive liikre of .: dp/ ys/ wi/hin llc phe or /he dlosd pikinj Mhen Closed /Le guk eon /sininw/ would prec/ude re/ arse s/g/84 dyslem ca/s!de j'inside Amend. 10 dottleinmeri/ d/mospdere /c lAe knvirQ 6.2.6-6

- m ~ m p n m

. TABLE 6.2.6-1 (SHEET l- OF 2)

CONTAINMENT PENETRATIONS NOT DRAINED DURING TYPE A TESTING Pene t ra t ion petember tal pejc_ ring f ont'l h Irication The steam generator secondary side is 1-Is. 7-10, 118, 11C, Steam generator secondary side (main stea lowdown, r, and considered an extension of the con-tainment. The system is not part g

128 12C. sampling, reedw.

18-21, 101-108: auxiliary feedwater) of the reactor coolant presstere bounda ry and does not open directly to the containment atmosphere during normal and post-accident conditions.

30, 31 33 Sarety injection lino the system is normally filled with water troe refueling water storage tank and g operating tender post-accident

' conditions.

32 Boron injection line The system is normally filled with water (high head sarcty f rom the cha rg i ng pump d i scha rge and ,

injretion) operating under post-accident

~

- - - - - - - conditions. _ _ _ _ _ _

10 0

38s, 3 5 containment spr: y supply The system is norma Iy ri f led with water 'o rrom refueling water storage tank and I i J

operating under post-accident M i s

" N eonditions.

36-39 Residual heat removal and The system is normally rilled wth water e os containment spray ptemp f rom refueling water storage tank and suction from containment operating under post-accident conditions.

emergency sump Af ter an accident the static head between emergency sump and valve provides water seal to prevent leakage or containment atmosphere.

56, 57. 58 Residual heat removal The system is normally filled with water discharge to reactor coolant f rom the rerteeling water storage tank system and operating under post-accident conditions, y 59, 60 Residual heat removal sortion from hot leg The system is closed outside containment and constructed to ASME Ill. Class 2,

, and Seismic Category I standards.

y On I

H f

I I

%D  :

ss ..

03  !

u

E

^ ^ p e q e ,

^ D I !, .

q

't l

TABLE 6.2.6-1 (SHEET 2 OF 2) ,

.l '

'! Penatraeion . DescripM on ,

Justirigajig,g j IttaQen i Nuclear service cooling water The system is a closed system inside

l - 's a 3 46, 91-96 '

-supply rod return containment per.COC 57 and thus not a

.j- potentia l at><.opsheric leak pa th. The

, system is it.9.aally rilled with water and _ -

operating. post-accident. Tle system is also raquired to cool the containment atmosphere during the Type A tests.

RCP sh*f whter supply ,

The system is closed outside containment

, $1-5ts and constrtK:ttd to ASME III. Class 2 and S fsmic Lete?ary 1 standards. The .

system is fil!?d with water during'all j

modes or riant operation (normal and post-acc!sent) by the charging pumps.

The system is tilled with liquid and l

13C,.67C. Containment pressure- , designed ic settsry the requi rements or 1 ,

'69C. 70C, detectors i Ve4.' Also the system 1 Regulatory Wiccdside and outside the 10 I 73C, 85C is closed t:oth i O containment, g ,

Reactor vassel ter levet The system la filled Nith liquid and $

I 14A, 149, inst rter-ta t f ort designed to S&tisfy the requirements of m

~ 14C, 76A, s Regulatory Guide'fh44', Also the system w l 768, 76C is closed both i kside and outside, e4, a

(

• G

" l.141 i i

. u 1

I 5

i en -

D O,

4 y a. See table 6.2.4-1 for further description.

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__ __ IF cI .39

. VEGP-FSAR-Q k- i Question 480.33 FSAR paragraph 6.2.6.3 states.that type C testing of the safety injection lines, containment spray lines, and long term l recirculation lines will not be done on the basis that these lines are water sealed. Additional justification is needed for the elimination of type C tests (note that table 6.2.4-1 indicates type C testing for the spray lines):

8 A. For each line, discuss and justify that a sufficient

( water inventory will be available for at least 30 days following a loss-of-coolant accident (LOCA).

B. For each line, discuss your plans for hydrostatically testing the valves to show that water leakage from the isolation valves is compatible with the 30-day inventory requirement. The leakage limits for these valves should be included in the plant Technical Specifications.

C. FSAR paragraph 6.2.6.3 states that the isolation valves, in the charging line of the chemical and volume control system are type C tested using water. Type C testing using water as the test fluid is permissible only if it k- can be shown that parts A and B above are satisfied.

Response n d . M is t o f

• Cl03ed dySlC"r Cul Side.

onbnmmen t The lines which penetrat: ::nt: inn;nY :nd*are required to perform a safeguard function following an accident are not Type C tested. Each of these lines is equipped with isolation valves that can be actuated by the operator from the control room.

Lines which fall into this category are:

A. Safety injection pump discharge lines (penetrations 30, 9 31, and 33).

(

B. Residual heat removal pump discharge lines (penetrations 56, 57, and 58).

l C. Centrifugal charging pump cold leg injection path

{ (penetration 32).

l \

o_-

D. Containment spray pump discharge lines (penetrations 34 and 35).

(

Amend. 8 7/84 Q480.33-1 Amend. 9 8/84 l

l

' " ~ ~

tr s*

VEGP-FSAR-Q .

D.

.E'. Containment emergency sump lines to the residual heat removal and containment spray pumps suction  ;

(penetrations 36 through 39).

Residual heat removal pump hot leg suction lines (penetrations 59 and 60).

Jens ferm emersem an"$ red 4)satety injection, residual heat removal, . .n t ;n- ;r. eprayf.and -

j high head safety injection portions of the chemical and volume control system are closed systems outside the containment, designed and constructed to ASME III, Class 2, and Seismic Category 1 requirements;/Theabovesystemsareoperatedand inspected during normal ' plant operation to ensure -=n thathathew integrity is maintained. -A aiepl actiec failure l

ac c c--ne ted- -

j __

)

[The reactor coolant pump seal water injection lines {

9 (penetrations 51 through S4) are not required to perform a safety function during an accident. However, due to the \

sensitive nature of the seals, it is highly desirable to providO seal flow at all times. The charging pumps are used for high head safety injection, and flow will be provided by these pumps through the seal injection lines following an accident. In addition, the system is designed, constructed, and maintained to qualify as a closed system outside the containment. Each line i

is equipped with a remote manual containment isolation valve j which the operator can close if Yequired.

The isolation valves in the above lines are either normally open atthetimeofanaccidentorareopenedatsometimeafterthem) accident to effect immediate and long term core cooling. Y.@NSE4.7

-hrehr the liner er: continuoucly "ater filled duri" L _

c%mm :---- - vlius system veerstien, either frcr ther~en:f-

-+ er-fra= +k%;.tsin :n+

s r;fueli-- : + -tm _g:

sumps and-:: cuch de net pivvide a credible path fer leekag: #

h* = 4 nm ne = +m n pherei*--

ks been I

Table 6.2.4-1 and subsection 6.2.6 pill 5"~ revised iu Ammndment

40"to reflect the above discussiony- and /, /ndica/e. f%d /Ae.

een/dnment spray famp disebaffe. kne. Yalves{ pena /,,/jon, g,,) 35) are. hpe C lea /ed.

bak PA Ucy do ns/ consM./e a. pv/enL/ conhumes/ a/essphere de/ Ire S/ure ol +

d""") or Al*% an accalenf w A a. 6 f le.

dys/m fomponal ~

Q480.33-2 Amend. 9 8/84

- . _a r e ..

W . _ . ., . . . . . . .

. . _ _ . . . .. ..ZNSE2.T . . .To _ 9'80. 33 i-b 1 wh;ch . .are closed /Wll</$ or cAosed s/ p,/ime ,

44 I' I .

. GAking.a. /oss o/- coolsn/ sccidenf are resi/ilmul./o effec / I 1 Vl . . .

. proper .dysJem . opera tion and tro f fa ellec/ a. barrier sj iral SAca/d /Ae yalaes leak

. release o/ conbinmes/ s/xarpiere.

e/osed, /Ae //uid 6ta/ A/Ain /Je pr;oe or s/gA/]/r vden ou/ side 1(de conkihenen/ ajoulof fAe c/ased pipg sysfern lo /de enviroin.

preelude . relase es' conkkmen/ dmdere The .yake dfems are no/ s- po/en// / con /sinmen/ s/mosphere warn /Je values sre e/osd 4r one of /Aa 4//. wig

^

lea k pa/4 rensans.

- tA' e valve s/ern is capped:

is ene/asal ~ in sn enqu/a/ian vessel

.Bre. value th<. valec. 6/em /eakage. is routed i. h. reyle Ao/ dup isnk which has 4. Ar '

is con /4 mas.fy venfilaied wiM the disp saf* ym. /y relald pipig pene/rs. lion exhaus/ mslem (Gee Ajare f 3.+'-2 and paragapA 1 ts.2)

(

is

- "- {The space aber.+be a;,ps,aym ne pressure

...a--[yy{:-  :

. gel. of .HeQoundary = .cmd. glive

-> L . . .. - .. .

.g; J.

. ._...p... . . . -

.7 4

. _. . 4 - _.

. . . ._. L.. . _.._.....7

I - -. -. . - .

}

s. .

l' l 1 I

O VEGP-FSAR-Q

---*-4----'*'

B. Safety injection, residual heat removal,

7 y? and ong term recirculation lines (penetrations  !

30 throug , and 56 through 60). See the response to question 4 0.33. g C. Nuclear service cooling water (penetrations 43 through >

46, and 91 through 98). Nuclear service cooling water to and from the containment fan coolers is a closed '

system inside containment. The nuclear service cooling water system inside containment is designed and constructed to ASME III, Class 2, and Seismic Category I requirements. No single active failure will provide /

a leakage path for the post-loss-of-coolant accident containment atmosphere.

D. Reactor coolant pump seal water injection (penetrations 51, 52, 53, and 54). See the response to question 480.33.

E. Containment pressure and reactor vessel level instrumentation (penetrations 13C, 14A, 14B, 14C, SSA, ' 9 67C, 69C, 70C, 71C, 76A, 76B, 76C, and 85C). These penetrations are designed to satisfy the requirements 14 of Regulatory Guide 1.141 These lines have no isolation valves and rely on a closed system both inside and outside the containment to preclude the release of containment atmosphere to the environment.

The integrity of these penetrations is verified during '

the periodic Type A tests.

F. Integrated leak rate test connections (penetrations 64A, 64B, 68, and 87). Following the completion of the integrated leak rate test (i.e., Type A test) which is i conducted using these penetrations, these penetrations are Type B tested in accordance with 10 CFR 50, Appendix J.

l G. Equipment hatch, personnel locks, and emergency doors.

These penetrations are Type B tested in accordance with ,

10 CFR 50, Appendix J. ,

H. Transfer tube (penetration 89). This penetration is Type B tested in accordance with 10 CFR 50, Appendix J.

l Amend. 9 8/84 Q480.24-2 Amend. 14 2/85

.- .. .. . . . - . . . . . . . - - - . - . . - . . . . .