ML20113D385

Responds to 850322 Request for Addl Info Re Vol Reduction Sys Per Draft SER Open Item 89.Further Analyses Made to Facilitate Assessment of Impact of Possible Incidents & Accidents in Realistic Manner

ML20113D385
Person / Time
Site:	Vogtle
Issue date:	04/04/1985
From:	Bailey J GEORGIA POWER CO.
To:	Adensam E Office of Nuclear Reactor Regulation
References
GN-577, NUDOCS 8504150264
Download: ML20113D385 (14)

Text

._ _

- Georgia Power Company Routs 2. Box 299A

• Waynisboro Georgim 30830

- Telephone 404 554-9961 404 724 8114-Southern Company Services,Inc.

Post Office Box 2625 Birmingham. Alabama 35202 Telephone 205 870 6011 Vogtle Proj.ect April 4, 1985 Director of Nuclear Reactor Regulation File: X3BC35 Attention: Ms. Elinor G. Adensam, Chief Eog: GN-577 Licensing'-Branch #4 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 V0GTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 REQUEST FOR SUPPLEMENTAL INFORMATION DSER OPEN ITEM 89 - VOLUMN REDUCTION SYSTEM (VRS)

Dear Mr. Denton:

Attached for the review of your staff are additional items concerning the VRS. These items were asked for by your staff in a meeting on March 22, 1985 and includes the following information.

o VRS Accident Evaluation e VRS Conformance to R.G. 1.140 o Comparison of Table 7.5.2-1 and Table 11.5.5-2 e Concerns on VRS startup and operation If your staff requires any additional information, please do not hesitate to contact me.

Sincerely, J. A. Bailey Project Licensing Manager JAB /sm Enclosure xc: D. O. Foster R. A. Thomas G. F. Trowbridge, Esquire J. E. Joiner, Esquire C. A. Stangler L. Fowler 8504150264 850404 M. A. Miller PDR ADOCK 05000424 0gf L.~T. Cucwa E PDR i g

G. Bockhold, Jr. I j

0145m

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VRS Accident Evaluation VEGP-OLSER-7 7.0 ENVIRONMENTAL EFFECTS OF ACCIDENTS 7

7.1 STATION ACCIDENT INVOLVING RADIOACTIVITY 7.

1.1 INTRODUCTION

- Accidents which cause concern from the environmental protection standpoint are those which might result in an uncontrolled release of radioactive materials to the environment. Numerous barriers and features are provided which guard against accidental or uncontrolled releases of radioactive materials from the plant. These barriers are: the sealed metal cladding tubes which contain the fuel pellets; the reactor coolant  ;

system which encloses the reactor; and the containment which houses the reactor coolant system. Additional protection of the public is provided by safety features which control the release of radioactivity in the event of an accident and by the site location which further reduces the potential effects to the general.public of an accidental release of radioactivity. ,

Various postulated-incidents and all accidents except severe 3 accidents have been analyzed and reported in detail in the Final Safety Analysis Report (FSAR) for the VEGP. These analyses demonstrate that the pisnt can be operated safely and '

that maximum radiation exposures from credible accidents would be within the guidelines of 10 CFR 100. Using very conservative calculations and assumptions, the doses calculated are in excess of what would be-realistically encountared.

Appendix 7A discusses the probalistic assessment of severe

! accidents heretofore called Class 9.0 accidents.

To facilitate the assessment of the impact of possible incidents and accidents in a realistic manner, and therefore to allow a judgment as to the potential environmental risk i

inherent to the operation of VEGP, further analyses have been made. Compared to the FSAR analyses, the environmental risk l analyses are intended to be more realistic. For example, realistic values have been assigned to such parameters as filter efficiencies and atmospheric diffusion.

7.1.2 ANALYSIS OF ENVIRONMENTAL EFFECTS OF ACCIDENTS l'

A variety of accidents and incidents have been analyzed l

covering a wide range of severity to facilitate a realistic assessment of environmental risk. Table 7.1-1 summarinas the events which were considered. These represent a spectrum of events from relatively minor to the most severe which could credibly be postulated. The classification follows that of

' Regulatory Guide 4.2, Revision 2, Preparation of Environmental 7.1-1

VEGP-OLSER-7 Reports for Nuclear Power Plants ( l'J7 6 ) . Calculated results of these events are shown in table 7.1-2 in terms of exclusion area boundary and integrated population doses. Details of the parameters used for each accident are included in the discussion of that event.

7.1.3 DOSE CALCULATION METHODOLOGY The radiological impacts of the postulated events are evaluated in terms of the radiation doses delivered to individuals and to the population as a whole. Whole body doses due to external exposure and thyroid doses due to inhalation are calculated for an individual at the exclusion area boundary and for the population within 50 miles of the plant site. The calculated exposures are limited to whole body and thyroid gland because these are the critical exposures for the radionuclides of potential concern.

The doses were calculated using the GASPAR computer program described in NUREG 0597, based upon the methodology of' Regulatory Guide 1.109, Revision 1, for atmospheric releases.

Population doses were. computed based upon the projected

~

population for the year 2007. The population projections were developed from the 1980 census and are described in FSAR subsection 2.1.3.

The dispersion factors used in the calculations were developed from 3 years of actual meteorological data for the VEGP site and are presented in FSAR subsection 2.3.5 For the purpose of this analysis, all releases are assumed to be ground level releases into the building wake.

7.1.4 TRIVIAL INCIDENTS (CLASS 1.0)

Pursuant to Nuclear Regulatory Commission (NRC) Regulatory Guide 4.2, Revision 2, Class 1.0 incidents have not been considered because of their trivial consequences.

7.1.5 SMALL RELEASES CUTSIDE CONTAINMENT (CLASS 2.0)

Pipes, valves, and flanges of systems containing fluids or gases with potentially significant radioactive concentrations

are designed, fabricated, and erected to minimize leakages that may occur during normal plant operations.

Although constructed with the intention of having no leakage, wear- and use-related activities can cause small leakage source terms. These low level releases are evaluated as routine releases and are included in the plant release source terms 7.1-2 ._ _ _ _ _ _

1 VEGP-OLSER-7

. ossibility of a small crack or diaphragm leakage -

re ng low level leak rates is given primary cons ation in the d .of the system and components. The rec .oldup tanks ar subject to high pressures or unus 's t r e s s e s .

Because of e factors, the possibility of ,ilure of the recycle holdu k is considered small.

In the unlikely ev hat a release o quid radioactive I wastes does occur fro recycle p tank, the spilled I liquid collects in a wat ht r _and a high water level l alarm will be activated in rol room.

l In view of the above discus , ossibility of an accident

of this type occurring i .

sidere. all.

7.1.6.3.4 Radiolo 1 Effects N

Using the ass ons stated, the followinh dos ve been calculated:

Whole Body

\ Thv.

Exc on area boundary (mrem) 2.59 x 10-' l'19 ulation dose (man-rem) -- 8. 2 7 x 10 -' 3.01 x 10 2 7.1.6.4 Equiement Leakage or Malfunction of the Volume Reduction System (Class 3.3) 7.1.6.4.1 Description The postulated accident is defined as a leak or malfunction that results in the release of a portion of the inventory of the volume reduction system. The release is the volume reduction system's airborne dry product.

7.1.6.4.2 Calculation Assumptions i

A. The leakage is assumed to occur in the line from the fluid bed dryer to the Venturi scrubber.

B. The duration of the release is 24 h. The reiease rate is 100 ft'/ min, corresponding to the maximum leakage possible without automatic system shutdown.

C. A ground level release from the building is assumed.

7.1-5

VEGP-OLSER-7 D. The airborne radioactivity released via the cubicle heating, ventilation, and air conditioning (HVAC) is processed through a high efficiency particulate air (HEPA) filter and a charcoal filter prior to environmental release.

E. A filter efficiency of 99 percent is assumed for the charcoal filters.

F. Due to the HVAC filters and the relative radiological effects, only isotopes of iodine have been censidered.

Total activity released to the environment following the accident is included in table 7.1-3, part III.

7.1.6.4.3 Probability of Occurrence The likelihood c. the occurrence of this type of accident is small. The volume reduction system is equipped with an exhaust monitor to ensure that the parameters are within the prescribed limits. In addition, routine surveillance will note any upward trend and thereby limit the probability of occurrence.

7.1.6.4.4 Radiological Effects _

Using the assumptions stated, the following doses have been calculated:

Whole Body Thvroid Exclusion area boundary (mrem) 5.5 x 10** 4.1 x 10 "

Population dose (raan-rem) 1.56 x 10 ' 6.33 x 10 "

l l

T' .6.5.1 Description  !

This po accident is defined as an unspecifi Y that l initiates .-h4 . rupture of a waste gas .. . The l airborne raditantiva. ased from this

• ring the i accident is assuded to be dir - .o the environment via the plant vent.

t 7.1.6.5.2 Ca Assumptions l hundred percent of the average tank innteq l ,- been assumed to be released, as shown in tables..4s'-

i

• \

7.1

._. ___ ._ _ ___________._-6 _ . _ _ _ . _ _ _ . . _ _ _ _ - _ _ . _ . - . . _ _ _ . . _

VEGP-OLSER-7

/

part IV. This evalua/ tion is based on normal ope .ng v

conditions.

B. airborne rad oactivity released nto xiliary lding has be'en assumed to be rel' eased itered to nvironment. /

7.1.6.5.3 Prob . 4 ty o ccurrence The likelihood of a

/ advergent , waste decay tank rupture is considered small. Th dioac gas tored in the decay tanks wilL' consist of . on p educ+ ses, hydrogen, and nitrogen cover gas. The rdgen\y ' be added in the<various collection and holdup tank pr ude the possibili-ty of obtaining a flammable mixtur d. gen gas. Henc'e , a tank rupture resulting from iq'iti n f hydrqgen in the/ decay tank is considered remote. The/syst alsoNbe designed to appropriate industry (nd S ic gory \ component standards. In addition, aste p ssing ystem panel with associai d alarms, iso on valves, systgm surveillance will ensure that .he sibility of th type of accident is small. /

o Effects - /

7.1.6.5.4 Using the

/ ad'/g R

mptions stated, the following dos

/ .-

have been calculate Whole Body Th -

N Exc ion area boundary (mrem) 2 . 42 x 10 - 8 2.42 x

'\

ulation dose (man-rem) 2.74 x 10 2 , 274 x 10 -

l l 7.1.6.6 Ruoture of a Licuid Radwaste Holdue Tank (Class 3.5) l 7.1.6.6.1 Description This postulated accident is defined as an unspecified event that initiates the complete rupture of the tank containing the largest quantity of significant isotopes in the waste management system. This tank has been identified as a recycle holdup tank located in the auxiliary building. The airborne radioactivity released from this tank during the postulated accident is then vented to the environment via the plant vent.

l 7.1-7

VEGP-OLSER-7 7.1.6.6.2 Calculation Assumptions A. One hundred percent of the average inventory of a recycle holdup tank has been assumed to be released into the auxiliary building.

B. An iodine partition factor of 0.001 for air to water has been assumed.

C. The airborne radioactivity released has been assumed to be released unfiltered to the environment ( table 7.1-3, part V).

1 7.1.6.6.3 Probability of Occurrence The discussion concerning the remoteness of an equipment leakage or malfunction accident of a recycle holdup tank is equally applicable to a complete release accident. The possibility of a complete rupture or complete malfunction accident is therefore considered even less than that of the partial release accident described in paragraph 7.1.6.3.

7.1.6.6.4 Radiological Effects--

Using the assumptions stated, the following offsite doses have been calculated:

Whole Body Thyroid Exclusion area boundary (mrem) 2.08 x 10** 1.54 Population dose (man-rem) 1.03 x 10** 3.76 x 10~2 7.1.6.7 Rupture of the Volume Reduction System (Class 3.6) 7.1.6.7.1 Description This postulated accident is defined as an event that causes the complete rupture of the volume reduction system. The airborne radioactivity released from the system during the accident is assumed to be vented to the environment via the radwaste building vent.

t 7.1-8 l

VEGP-OLSER-7 7.1.6.7.2 Calculation Assumptions A. Isotopic source terms are assumed to be 100 percent of

~the iodine in suspended dust particles in the volume reduction system.

B Activity released is assumed to consist of the airborne solids in the system from the dryer bed to the Venturi scrubber.

C. The airborne radioactivity released has been assumed to have passed through a charcoal filter and the building's HEPA filter prior to being released to the environment.

D. The release is assumed to be a ground level release.

E. A filter efficiency of 99 percent for the charcoal filters has been assumed. Total activity released to the environment following the accident is included in table 7.1-3, part VI.

7.1.6.7.3 Probability of occurrence The likelihood of a rupture of tHn volume reduction system is small because a change in system pressure greater than 2 psi will cause the high/ low pressure switches to shut the system down automatically. Also, the system has a maximum operating pressure of 10 psig and hydrostatic test pressures greater than 25 psig. This will further reduce the probability of this fype of an accident. In addition, the Volume Reduction System instrumentation and controls maintain process parameters within limits which ensure safe system operation.  ;

7.1.6.7.4 Radiological Effects Using the assumptions stated, the following offsite doses have been calculated.

Whole Body Thyroid Exclusion area boundary (mrem) 3.27 x 10 2 . 42 x 10 *

• Population dose (man-rem) 9.2 x 10 3.74 x 10**

I 7.1-9

VEGP-OLSER-7 TABLE 7.1-3 (SHEET 1 OF 8)

ACTIVITY RELEASED TO THE ENVIRONMENT Isotope Activity Released (C1)

Part I - Following a Waste Gas Decay Tank Equipment Leakage or Malfunction Accident Kr-83m 4.0 x 10-2 2 Kr-85 1.7 x 10 Kr-85m 6.8 x 10'1 Kr-87 6.4 x 10 2 Kr-88 6.9 x 10**

Kr-89 1.4 x 10*'

Xe-131m 1.5 Xe-133 2.2 x 102 Xe-133m 2.8 Xe-135 3.5 Xe-135m 7 . 7 x 10 -'

Xe-137 3.7 x 10*'

Xe-138 2.3 x 10-8 Part II - Following a Liquid Waste Tank Leakage or Malfunction Accident I-130 3.3 x 10-8 I-131 6.2 x 10-8 I-132 3. 6 x 10 -5 I-133 9.3 x 10

I-134 5.3 x 10-'

I-135 1.6 x 10 '

sm Part III - Following a Volume Reduction Leakage or Malfunction Accident I-130 5.52 x 10 '

I-131 1.56 x 10*'

I-132 3.12 x 10**

I-133 3.29 x 10

I-134 8.59 x 10 '

I-135 1.86 x 10 '

VEGP-OLSER-7 TABLE 7.1-3 (SHEET 2 0F 8) ,

Isotope Activity Released (Ci)

Part IV - Following a Rupture of a Waste Gas Decay Tank Kr-83m 1.55 x 10 1 Kr-85 2.69 x 10 1 Kr-85m 2.65 Kr-87 2.52 x 10 1 Kr-88 2.7 Kr-89 5.52 x 10 5 Xe-131m 1.2 x 10 1 Xe-133 1.58 x 108 Xe-133m 1.58 x 101 Xe-135 1.58 x 10 1 Xe-135m 3.0 x 10 8 Xe-137 1.45 x 10**

Xe-138 8.82 x 10 '

> Part V - Following a Rupture of a Liquid Waste Holdup Tank I-130 1.3 x 10 '

I-131 2.5 x 10**

-I-132 1.4 x 10 '

I-133 3.7 x 10 '

I-134 2.1 x 10

I-135 6. 4 x 10 -*

Part VI - Following a Rupture of 7 the Volume Reduction System I-130 3.6 x 10 '

I-131 1.02 x 10 '

I-132 2.03 x 10 7 I-133 2.14 x 10 '

I-134 5.6 x 10"'

I-135 1.21 x 10-'

i

VRS R.G. 1.140 Evaluation Operation and Testing of Off-Gas System with Regulatory Guide 1.140 Regulatory Guide 1.140 requires that the HEPAs and charcoal adsorber be procured and tested prior to instal-lation in accordance with ANSI N509. "Ehis should not be a problem since we have committed to similar th6egs in FSAR section 9.4. *# N ge,*,,*4 h

Section C.5 of Regulatory Guide 1.140 requires four tests to be done upon initial installation.

a. Visual Inspection

b. Airflow distribution for HEPA filters (bypass leakage)

c. DOP tests for HEPA filters

d. Adsorber leak testing (Air-Distribution and Tracer Tests) 63 rs **

The 4_S'- f S filters were designed with sample and injection ports,so they have the ability to carry out the required tests. Our filters are contained in a pressure vessel with no space between the HEPAs and the charcoal.

The visual inspection outlined in ANSI N510 can be accomplished during installation. An overall pressure drop test can be accomplished after installation and then compared to the design data. The design of these filters 2204t

is that of a process filter. The VR off-gas system is hard piped before and after the filter vessels. The size flow of the piping is not amendable to accurate ^ distribution J

readings as re av red fee +h, fja ,g;,;,;y,q,,, 4,,4, The installation of the filter assembly follows a specific procedure that minimizes the potential for the HEPAs and adsorber to be damaged.

The DOP and tracer tests cannot be accomplished on an

< individual or overall basis at this time.

Section C.5 also requires in-place HEPA DOP and adsorber leak testing at intervals of every 18 months. The expected change-out frequency of the filters is every 6 months at which time the on-line filter is valved out of service and allowed to decay. At that time the idle filter is valved into service and used. Because of the

+ e s4s changaout frequency, the 18 month intervaleAare not applicable.

FSAR questions 460.06 and 460.12 deal with the filters and compliance with Regulatory Guide 1.140, 2204t

r

. Respouxes to NRC Concerns on Table 7.5.2-1 & R.G. 1.97

1. Condenser air ejector has a range of 5x10-7 to 105 in Table 11.5.5-2 and this range has been included in update of 7.5.2-1* in Amendment 16. Therefore both ranges are identical.

2. Plant vent - Table 7.5.2-1* inadvertently did not include the exponents for the ranges for this monitor. The Table (7.5.2-1) will be updated in a future amendment to be made consistent with Table 11.5.5-2. The range of this monitor is 10-6 to 104,

3. Steamline radiation monitor - both Tables 7.5.2-1* and 11.5.5-2 are consistent.

4. Radwaste building monitor - this monitor is not in Table 7.5.2-1 because it is Category 3, and thus is not used as a key or backup variable for accident mitigation.

5. SG liquid - this monitor does not have to be included in Table 7.5.2-1 since it is isolated on any signal that gives an auxiliary feedwater start, and therefore cannot be used for accident diagnosis.

6. Plant vent /radwaste building samplers - a range for these passive samplers is inappropriate. The samplers will be analyzed in the lab to determine activity.

• As shown in GN-548, dated 3/15/85, DSER Open Item 62 - R.G. 197, Rev. 2 0143m

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Responces to NRC Concerns on VRS System l

Q. #1 Will instrumentation be calibrated prior to initial startup of VRS?

A. Yes. Instrumentation will be calibrated according to recommended manufacturer guidelines. Calibration will be performed by qualified personnel using VEGP approved instrument calibration procedures.

Q. #2 Will the feed to the VRS exceed the 0.3% sulfur and 1% halogenated plastic limits?

No, we will prevent the feed from exceeding the limits by segregating out the high sulfur and halogenated plastics and not feeding them into the VRS. In addition, the purchasing department will be pro-hibited from buying high sulfur and halogenated plastics.

,