ML19323A822

Submits Partial Response to IE Bulletin 79-01B, Environ Qualification of Class IE Equipment. Nuclear Reactor Is Cooled by Gas,Not Water.Submergence Is Not Deemed to Be Problem.Equipment Lists & Equipment Design Review Encl

ML19323A822
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/18/1980
From:	Swart F PUBLIC SERVICE CO. OF COLORADO
To:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML19323A823	List: A14212, Environ Temps in Vicinity of Rupture Point of Steam Lines for Equipment Qualification. Description of Equipment Location in Facility Encl ML19323A822
References
IEB-79-01B, IEB-79-1B, NUDOCS 8005060064
Download: ML19323A822 (24)

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March 18, 1980 Fort St.<Vrain Unit No. 1 P-80051 Mr. Karl V. Seyfrit, Director Nuclear Regulatory Commission Region IV Office of Inspection and Enforcement 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76012

Subject:

Environmental Qualification of Class IE Equipment

References:

IE Bulletin 79-01B P-80037, March 4, 1980 Swart to Seyfrit j

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Dear Mr. Seyfrit:

As indicated in our letter of March 4,1980 (P-80037), PSC is hereby submitting a partial response to IE-79-018. Along with the submittal, we are summarizing the actions that are being taken as a result of IE Bulletin 79-018. Justification is provided for not responding in areas that we feel do not apply to an HTGR. Also provided is the current status of our response and the schedule for its completion.

General As previously submitted to the NRC (Attachment D to P-77137, Millen to Denise, June 15, 1977, enclosed as Attachment D), the double-ended rupture of a cold reheat pipe in the Reactor Building produces the most severe environmental conditions in that building.

Similarly, the double-ended rupture of a hot reheat pipe produces the most severe environmental conditions in the Turbine Building.

For the conditions resulting from a double-ended rupture of a hot or cold reheat pipe, components were qualified to the distance nearest the opposite loop steamline or to the nearest steamline if the component must function even if its own loop fails. If the distance was equal to or greater than 20 feet, the conditions of the 20 foot curve for the reactor or turbine buildings was used to qualify the components.

'Mr. Karl V. Seyfrit, Director March 18, 1980 Page 2 Partial Submittal (Attachment A)

Our submittal consists of a listing of Class IE electric equipment within the " accident zone" that is required to electrically function under accident conditions to provide safe shutdown cooling.

The submittal is formatted similarly to enclosures #1 and #2 of IE-79-018.

Equipment locations are given in the column entitled " SRB LOCATI0ti", which is defined in Attachment E.

The submittal that is similar to enclosure #2 is a tabulation of " tagged items." The term " tagged items" refers to equipment, instruments or components that are identified by a specific number.

The numbers are alpha numeric in nature and provide the following infonnation:

1.

The alpha portion identifies the type of component, fe; as, P= Pump, HS= Hand Switch, etc.

See Attachment B for a complete list of the Alpha prefixes.

2.

The numeric portion identifies the system involved. Specit ically, the first two digits identify the system involved. See Attachment B for a complete list of system numbers and names.

The submittal that is similar to enclosure #1 consists of "untagged" or "subtier" components. These subtier components are the many items such as relays, switches and other control components that are required as a part of the control loop or circuitry to make the tagged (equipment) item function.

These types of items do not carry any specific identification other than the manufacturer's model or part number and are generally not shown on the plant P&I drawings.

Suninary o_f_ Actions The following areas are being investigated in relation to our response to IE-79-01B:

1.

The environmental test records for the possible inclusion of additional subtier items.

2.

The computer programming required to format our final response similar to enclosure #3 of IE-79-018.

3.

Equipment suppliers are being contacted regarding aging.

4.

The Emergency Procedures are being reviewed.

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' Mr. Karl V. Seyfrit, Director March 18,1980 Page 3 Areas Not Applicable to an HTGR Pressure:

The FSV HTGR does not have a containment building, therefore, there is no storage of blowdown steam and thus no ambient pressure buildup.

The reactor and turbine buildings are both vented.

Therefore, pressure transtents resulting from a high energy 1ine break will be very 1ocalized and short-term in nature.

Furti1er details about steamline rupture analysis at FSV may be found in Attachment D to P-77137, dated June 15, 1977.

Relative Humidity:

For the same reasons as discussed above under the pressure heading, Relative Humidity is not a problem at FSV after a high energy line break.

Chemical Spray:

No chemical sprays are utilized at FSV for cooling.

Radiation:

There are no radiological concerns directly associated with a high energy line break at FSV.

That is, the process fluid (steam or feedwater) is not, contaminated.

To postulate a radiological incident DBA #1 " Permanent Loss of Forced Circulation" and DBA #2 " Rapid Depressurization/ Blowdown" were considered.

DBA #1 provides the worst case radiological conditions, but the overall radiological concerns are minimal.

Complete details of this accident may be found in Section 2.1.6.b of P-79312 (Swart to Varga) dated December 28, 1979, enclosed as Attachment C.

In summary, the peak doses in the Reactor Building following DBA #1 are as follows:

Peak Gamma 180 Day Accumulated Location Dose Rate Time of Peak Dose (Rem)

Reactor Building 1.4 R/hr 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 400 (above Refueling Floor)

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' Mr. Karl V. Seyfrit, Director March 18,1980 i

Page 4 In conclusion, the reactor building will be accessible for short-term operations following such an accident.

The accumulated doses indicated above would have no operational effect on the Reactor Building equipment.

Submergence The nuclear reactor at Fort St. Vrain is cooled by gas and not water.

Shutdown of the reactor is accomplished by control rod insertion.

Emergency shutdown is accomplished by pressurized shutdown hoppers that drop boron balls into the reactor. Water is not used for shutdown or emergency core spray of the reactor in an HTGR.

Venting of the reactor cooling quench water and/or primary coolant water into the containment sump is not applicable for Fort St. Vrain.

Therefore, submergence is not deemed to be a problem at Fort St. Vrain.

Schedule Component evaluation worksheets similar to IE-79-018 Enclosure 3 will be submitted in a preliminary form within approximately 2 weeks, along with any revisions to the master list.

After the above is completed, revised versions of the master list and component worksheets will be supplied on a weekly basis if revisions are required.

The response to this bulletin and the schedule for its completion is based upon the use of the firewater system as the source of motive power for the helium circulators and water to cool a steam generator as described in the FSV FSAR Section 14.4.

This mode of reactor cooling utilizes only seismically and environmentally qualified equipment components and systems. With this in mind, there should be no problem in r;>eting the 90 day response deadline for IE-79-018.

If the review of the FSV Dnergency Procedures or any other item has a major impact on the schedule the staff will be so advised.

Very truly yours,

'M Frederic E. Swart, Manager Nuclear Project Departnent FES/ MEN:pa Attachments

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ATTACHMENT B Page 1 ALPHA PREFIXES Designation D_e scription A

Absorbers, Traps and Demineralizers C

Compressors, Blowers, Vacuum Pumps, Fans Including Drives E

Exchangers, Cooling Towers F

Filters, Strainers, and Dryers I

Instrument and/or Control Racks and Panels N

Electrical Power / Control Cabinets P

Pumps and Drives S

Special Packaged Items T

Tanks and Vessels FV Flow Valve FT Flow Transmitter HS Hand Switch HV Hand Valve LS Level Switch LT Level Transmitter LV Level Valve PS Pressure Switch PT Pressure Transmitter PV Pressure Valve SV Speed Valve TS Temperature Switch XE Special Element (Steamline Rupture Sensor)

ZS Position Switch FIS Flow Indicating Switch FSL Flow Switch Low HS1 Hand Solencia Valve LSH Level Switch High LSL Level Switch Low LSV Level Solenoid Valve PDT Pressure Differential Transmitter PSH Pressure Switch High PSL Pressure Switch Low TSH Temperature Switch High XEP Special Electrical Pneumatic Transducer PDIS Pressure Differential Indicating Switch PDSH Pressure Differential Switch High i

l

ATTACHMENT B Page 2 SYSTEM NUiBERS System Description 11 Reactor Vessel and Internal Components 21 Primary Coolant System 22 Secondary Coolant System 23 Helium Purification System 31 Feedwater and Condensate 42 Service Water System 46 Reactor Plant Cooling Water System 82 Instrument and Service Air 91 Piping-Hydraulic 011 System 92 Electrical-Including Switchgear and Standby Diesel Generator 93 Control and Instrumentation A

l l

l

ATTACHMENT C Section 2.1.6.b -- Desien Review of Plant Shieldinc and Environmental Qualification of Scuioment for Soaces/ Systems Which May Be U3ed In Post-Accident Ooerations PSC December 12,1979 (P-79299) REPLY:

"PSC will perform the radiation protection design reviews required by Section 2.1.6.b, utilizing the source terms recommended in Regulatory Guides 1.3,1.4, and 1.7, and will submit the results of the review to the NRC by January 1,1980.

Where doses received are in excess of GDC 19 guidelines, PSC will take those steps necessary to permit post-accident operations in vital areas.

Any required modifications will be completed by January 1,1981."

PSC December 27,1979 (P-79312) SUBMITTAL:

Tne assessment of post-accident operitor actions in vital areas at Fort St. Vrain (FSV) indicates that doses received from a hypothetical FSV accident scenario will not be in excess of the GDC 19 guidelines for the duration of the accident, provided the FSV reactor plant exhaust filters are adequately shielded.

PSC hereby commits to providing necessary shielding modifications to the FSV reactor plant exhaust filters by January 1,1981 to permit operator access to vital areas under accident conditions.

The. hypothetical Fort St. Vrain (FSV) accident scenario consists of the FSV Design Nis Accident (DBA) #1 ccmbined with successive PCRV primary coolant skage after depressurization.

For clarification, the DBA #1 and h. I leakage scenarios are explained below.

DISCUSSION:

To obtain a post-accident release of radioactivity equivalent to that described in Regulatory Guides 1.3,1.4 and 1.7 requires a permanent loss of all forced circulation for the FSV HTGR.

This specific accident was identified as DBA #1 in FSAR Section 14.10 and Appendix D.

These analyses performed by General Atomic Company at the time of licensing did not consider Regulatory Guides 1.3 and 1.4 source terms (i.e., the equivalent of the 50%

of the core radiciodine and 100% of the core noble gas inventory for release to the primary coolant) appropriate for the HTGR, However, because of past precedence by the then Atomic Energy Commission (AEC) of using the above source terms, offsite doses resulting from the postulated accident were calculated and presented in the previously mentioned FSAR sections using both the General Atomic Company release c.ssumptions and AEC TID-14844 release assumptions.

In both cases thre offsite doses are within 10CFR100 limits.

DBA di Descriotion:

A non-mechanistic loss of forced circulation is postulated from full power operation, where the reactor is scrammed by the plant protection system and all attempts to restore forced circulation using the multiple heat sinks, circulators and motive power for the circulators fail.

Because of the large heat sink provided by the graphite core, considerable time is available to initiate primary coolant depressurization and to restore forced circulation. The FSV FSAR specifies the time available to initiate depressurization to be 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, which was later amended by PSC letter P-77250 dated December 22,1977 to be 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The reduction in time was due to the capability of the helium purification system to process primary coolant during the planned blowdown of the clean primary coolant to the reactor building ventilation stack.

Tnus, the depressurization of the PCRV is initiated after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and completed 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> later (or 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> from the onset of the accident), at which time the PCRV has been depressurized to 5 psig.

The fuel is slow to heat up due to the large heat sink provided by the core graphite. A peak average active core temperature of 5400*F is reached about 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> after the onset of the accident.

At this temperature, the core structural integrity and geometry are not compromised since the vapori:ation temperature of graphite is 6900*F.

Peak activity released to the primary coolant, considering decay, is reached about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the accident.

Heat removal is provided by the liner cooling system in the redistribute mode which maximizes coolirg in the top head of the PCRV.

Leakage of primary ecolant from the PCRV is assumed to occur at a conservatively high leakage rate of 0.2% of the primary coolant inventory per day.

Offsite doses were calculated for a 6 month duration of the accident, but most of the offsite dose occurs in the first 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> of the accident, due to fission product decay.

Tne reactor building ventilstion system maintains continuous venting of the reactor building environm!nt at 1.5 volumes /hr during the entire period of the accident.

Primary Coolant Leakace Rate Durinc DBA #1:

Tne FSV FSAR DBA il ( Appendix 0, page 0.1-56) assumed an arbitrarily conservative and non-mec.1anistic estimate of PCRV leakage after the intentional depressurization by assuming that the liner has failed completely (or does not exist) and only concrete permeability centrols the leakage.

An internal 5 psi pressure differential was assumed which purportedly gave a PCRV leak rate of 8.33 x 10" fraction per hr (0.2%/ day).

Reference was made to Question IX.7 of Anendment No. 9 of the FSV FSAR for the calcualtion of tne pemeation rate for the FSV PCRV concrete under these conditions.

Examination of Question 0.2 revealed simply the conclusion that a 5 psi positive differential pressure led to 0.2%/ day and 2 psi positive i

dif**rential pressure led to 0.08%/ day.

Question IX.7 also did not provide details of the calculation of the 0.2%/ day rate.

However, considerable

4 detail and a derivation was provided for the analysis of leakage rate tests at high pressures.

The following equation was provided (egn.14 on page IX.7-8):

/p i W (1b/ day) = 1.13 x 10-5 in

+ 2.2 x 10-6 (p 2-- P 2 )

e 1

2 (egn.14)

Where

/P PCRV inside pressure in psig

=

6 P. = PCRV inside pressure in psig for which the net compressive stress in concrete = 0 2

A Face area of concrete, ft

=

X Concrete thickness, ft

=

Pz = Permeation or high side pressure, psia P2=

Ambient or low side pressure, psia Numerical values were inserted for Pz = 845 psig with the assumption that AP was approximately equal to P) in the following equation (egn.'S aa e

uma page):

b 1.13 x 10-5 in h + 9.1 x 10-7 9 0 ( 857.52 12.52) x

= 0.043 + 602 = 500 lb/ day (egn.15)

The first item to note is that the coefficient for the second (laminar flow) term is in error which is most likely a single error 1n transcribing from equation 14 to 15 since equation 13 has the 9.1 x 10~7 coefficient.

Equation 15 should read:

W = 1.13 x 10-5 9 0 8 2.'l+2.2x10-6 9000(857.52 12.52) x in

= 0.043 + 1445 = 1450 lb/ day (egn.15 revised)

The second item is that the LP/dPe term has been dropped in going frcm egn.14 to egn.15, which is significant if it is assumed that these equations are appropriate for evaluating the leak rate at Pz = 5 psig.

LEAX RATE Pressure P lb/ day

%/dav I

(psig) 5qn 14 15 15 Revisec 14 15 15 revisec Given App 0; Amend 9 Question 0.2 5

.0019

.13

.30

.001

.07

.17

.20 Amend 9 Question

, 0.2 2

.0003

.045

.107

.0001

.025

.059

.08 Since equation 14 is the appropriate equation, the 0.2%/ day leak rate is conservative by a factor of 200.

Furthermore, the only equation that comes close to the values given in the SAR is 15 Revised, that is, AP/AP. has been neglected which accounts for the factor of 200.

For purposes of plant shielding and equipment environmental evaluations, the historic 0.2%/ day is assumed to exist as an upper limit of all potential contaminated primary coolant leakage including permeability through the PCRV concrete. This is judged to be conservative since the primary coolant with ary significant activity is contained within the PCRV or helium purification components contained in wells within the PCRY.

Radionuclide Source Terms for DBA-1:

As previously stated, the fuel within the graphite core is slow to heatup during DBA#1.

Once it has reached the FSAR fuel particle coating failure temperature of 1725'C (3137'F) the fission products are assumed, for purposes of this shielding evaluation, to be realeased per the TID-14844 assumptions.

For release to the primary coolant within the PCRV, this is 100% of noble gases, 50% of the iodines and 1% others.

The total activity in uries contained in the primary coolant, assuming no leakage from the PCR/, as a function of lapsed time, is given in Table f.l.5.b-l.

Consistent with TIC-14844 release assumptions, 50% of the iodines plateout within the primary coolant system resulting in a depletion of the iodine to 25% of core inventory in the reactor building air.

Th9s, the total activity in curies in the reactor building, assuming the upper limit of 0.2%/ day leakage (which is being purged by the reactor building ventilation system a: the rate 1.5 volumes /hr), is given in Table 2.1.5.b-2.

TA8tE 2.1.6.L-1 HUREC-0578 STUDY TOTAI. ACTIV11Y (C1) PRESENT IN Tile PCRV ?RIHARY C001. ANT AT CIVEN EI.APSED TIHE (hours).

PCRV PitESSI BOUNDARY REHAINS 1HTACT.

TID-14844 Holdi4\\l.1ZATION FRACTIONS USED,100% HORI.E CASES, 501 10 DINE 1% OTilERS ELAPSED TIME (llours) 2 8

24 34 40 48 52 5 11 72 120 240 475 720 4320

}LIDE

-88 1.05104 2.89105 2.110:05 2.39804 5.89103 1.37103 5.50102 1.76102 7.04101 0

0 0

0 0

>-88 8.57103 2.79105 2.80805 2.66104 6.51-103 1.46103 6.07 02 1.89102 7.08101 0

0 0

0 0

r-95 3.15 01 6.66103 1.84105 2.57105 3.01:05 3.59105 3.69105 3.84105 4.18105 4.12105 3.88805 3.43105 3.02105 4.6010 r95 3.18101 6.74:03 1.87105 2.63105 3.09105 3.69105 3.80805 3.97105 4.35105 4.37105 4.31-105 4.12105 3.88105 8.9080

>131 1.33103 3.50105 6.111106 6.91106 7.33106 '7.88106 7.90106 7.93106 7.98106 7.57106 4.89106 2.07106 8.45105 0

F132 1.44103 2.34105 1.79106 6.09:05 5.64105 5.61105 3.68105 2.96105 2.72 05 1.76105 4.02:04 4.99103 5.46 02 0

l hl33 2.411 03 5.30105 6.44106 5.25106 4.70606 4.12:06 3.65106 3.05:06 2.04:06 4.84105 8.81:03 0

0 0

-133 5.25103 1.40:06 2.50107 2.78 07 2.94107 3.14107 3.14:07 3.12107 3.09:07 2.73107 1.41107 3.86106 9.90105 0

L135 1.98103 2.46105 1.40:06 5.49105 3.31-105 1.88805 1.25105 6.83 04 1.78104 2.94102 0

0 0

0 l

-135H 7.28102 8.34104 4.59105 1.72105 1.04105 5.97104 3.91-104 2.14-104 5.511:03 0

0 0

0 0

)-135 1.75103 5.43105 6.24806 3.86106 2.93106 2.11-106 1.62'106 1.08106 4.39805 1.81-104 0'

O 0

0 P140 5.44101 1.44104 2.58105 2.92 05 3.13105 3.39105 3.42105 3.45105 3.54105 3.57105 2.70105 1.56kO5 8.80104 0

'140 3.34 01 7.37103 2.01105 2.60805 2.93 05 3.36105 3.43:05 3.54105 3.75105 3.96105 3.10105 1.00105 1.01105 0

)-

s.

TABLE 2.1.6.b-2 IUREG-0578 STUDY TOTAL ACTIVITY (Cl) PRESENT IN lilE REACT 0ft BUILDING A1110SPilERE AT GIVEN ELAPSED TINE (liours).

PCRV LEAK RATE TO BUII. DING 0.2%/ DAY.

REACTOR llUII. DING VENTEI) AT 1.5 VOLttlES/flR.

TID-14844 NORHALIZED FRACTIONS USElt, 100% N0llLE CASES, 25% IODINE, 1% OTilERS ELAPSED TIME (flours)

ULIDE 2

11 24 34 40 48 52 5 11

___]2 120 240 475 720 4320 L88 3.77-01 1.31101 1.33101 1.32'100 3.22-01 7.10-02 3.00-02 9.23-03 3.3ft-03 0

0 0

0 0

8tl 3.55-01.1.37801 1.42101 1.411:00 3.58-01 7.77-02 3.34-02 1.02-0. 3.61-03 0

0 0

0 0

95 1.20-03 3.29-01 9.141100 1.40101 1.64801 1.97101 2.04:01 2.12:01 2.31101 2.29101 2.16:01 1.91 01 1.68401 2.56100 49 5 1.21-03 3.33-01 9.98400'1.43101 1.69101 2.02 01 2.10 01 2.19101 2.41-101 2.43401 2.39801 2.29801 2.16:01 4.94600

[31 2.52-02 11.64100 1.65102 1.90102 2.02:02,2.17102 2.19102 2.20402 2.21:02 2.10102 1.36102 5.76101 2.35801 0

L32 2.57-02 5.24600 4.24401 1.58101 1.46101 1.46101 1.05:01 8.46s00 7.75800 5.14100 1.30:00 1.61-01 1.77-02 0

33 4.68-02 1.30101 1.70102 1.44102 1.29 02 1.13102 1.01102 8.45101 5.65101 1.34101 2.45-01 0

0 0

3 133 1.93-01 6.94101 1.34803 1.54403 1.63103 1.75103 1.75803 1.75103 1.75103 1.52103 7.85102 2.14102 5.50s01 0

35 3.68-02 5.89100 3.60 8 01 1.51 f 01 9.05100 5,.09100 3.47100 1.89100 4.89-01 0

0 0

0 0

135H 3.14-02 7.71 00 8.59101 7.38101 5.53i01 3.53101 2.75101 1.81101 6.20:00 1.07-01 0

0 0

0 335 6.75-02 2.73101 3.40102 2.26802 1.72102 1.22102 9.56101 6.40101 2.56i01 1.01-100 0

0 0

0 40 2.06-03 7.12-01 1.38101 1.61101 1.72101 1.87101 1.89101 1.91101 1.96101 1.98:01 1.50801 8.67100 4.89400 0-

)40

1. 2 7-03 3.66-01 1.011101 1.43101 1.61101 1.85101 1.90 f 01 1.96101 2.08101 2.20101 1.72 4 01 9.98100 5.62 8 00 '0 l

l l

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' Radiation Levels Durine DBA-1:

Based upon TID-14844 source term release assumptions, the radiation levels were calculated in the reactor building ar.d the control room to determine the operator accessibility.

Details are described herein.

Assumotions In addition to the assumptions used in deriving the source terms, the fellowing assumptions were made for evaluating shielding adequacy:

1.

Credit was taken for a 50% depletion of the iodines due to plateout in the primary coolant system prior to release to the reactor building atmosphere.

i 2.-

All fission products were assumed to remain gasborne.

In other words, no plateout of fission products was contemplated.

3.

All the activities were uniformly distributed throughout the free space of the reactor building or the PCRV.

4.

The iodines and particulates removed by the reactor-building ventilation filters were deposited in any two of the three filters available.

5.

Only major shielding such as concrete walls was considered.

Reactor Buildinc To determine the accessibility of the reactor building during the course of DBA-1, the gamma dose rate in the reactor building was ca:culated as a function of elapsed time.

The contributing sources consist of the gasborne activity in the reactor building as a result of PCRV leakage, the primary coolant activity contained in the PCRV, and the buildup of iodines and particulates on the reactor building ventilation HEPA and charcoal adsorbers.

The contribution frcm the ventilation filters was not considered, as the filters will be properly shielded.

Two dose points were selected for the dose-rate calculation.

The first point is located at the center of the space above the refueling floor

(= 40 ft from the floor), and the second point is on the refueling floor directly above tne refueling penetration.

The PATH code described in FSAR Section 11.2.2.4 was utilized to perform the calculation.

Figure 1 shows the dose rate at the first dose point.

Essentially all the contributions come from the gasborne activity in the reactor building.

The activity in the PCRV is relatively insignificant to the first dose point, because of a large separation distance between the source and dose point.

Short-term access to the reactor building is possible.

The dose rate at the second dose point (i.e., on the refueling floor) is given in Figure 2.

The contributions from the reactor building and from the PCRV are individually represented, along with the total dose rate.

The contribution from the PCRV is due to the primary coolant activity present in the interspace below the primary closure for the control rod drive.

The maximum dose rate on the floor is 1.0 rem /hr, which is less than the peak dose rate of i.4 rem /hr at the first dose point.

Therefore, the refueling floor is accessible on a short-term basis. '

Control Room The dose rates in the control room include the contributions from the airborne activity in the reactor building atmosphere, and from the iodine and particulate activity accumulated in the plant ventilation filters.

The PATH code was used to determine the contribution from each source as a function of time into accident.

The dose point was located in the reactor engineer's office, as shown in Figure 3.

The results of the PATH calculations are shown in Figure 3 as a function of elapsed time.

It is apparent that the contribution from the airborne activity in the reactor building is relatively small or negligible as compared with that from the reactor building ventilation filters.

The dose rate reaches a peak of 700 mrem /hr about one month into accident. The important nuclides are Zr95, Nb95 and Lal40 accumulated in the filters.

The dose rate in the control room appears to be excessive for continuous manned access.

Adequate shielding will be provided for the ventilation filters so that the dose rate from the filters can be reduced to an acceptable level.

Summary Results The peak dose rates in the reactor building and control room are summarized below. Also indicated are the time at which the peak dose rate occurs following an accident, and the total dose accumulated over a period of 180 days from the initiation of the accident.

180 Day Peak Gamma Accumulated Location & Condition Dose Rate Time of Peak Dose (rem)

Reactor Building (above 1.4 R/hr 24 hrs.

400 refueling floor)

Control Room l

From Vent Filters (Unshielded) 700 mR/hr

= 720 nrs.

2400 From Reactor Building 3 mR/hr 24 hrs.

0.9

Conclusion The following conclusions are reached from tne review of shielding design adequacy for DBA-1 conditions and 710-14844 source term release assumptions:

1.

The reactor building ventilation filters will be adequately shielded to reduce the dosage contribution from the filters.

2.

Areas immediately outside the reactor building should be accessible only on a restricted basis, because of direct radiation from the activity in the reactor building.

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